ML20195H735
| ML20195H735 | |
| Person / Time | |
|---|---|
| Issue date: | 05/19/1999 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20195H628 | List: |
| References | |
| 0610, 610, NUDOCS 9906170119 | |
| Download: ML20195H735 (152) | |
Text
610_5_19.wpd NRC INSPECTION MANUAL PIP 8 MANUAL CHAPTER 0610*
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INSPECTION REPORTS DRAFT 1
9906170119 990520 PDR REVGP ERGNUMRC PDR Obl 76/ / j
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I 0610* DRAFT Issue Date: DRAFT 5/19/99
l INSPECTION REPORTS -
l Ighle of Contents.
E.agg a
0610-01 P U R PO S E.................................................... 1 1
0610-02 OBJ ECTI VE S................................................. 1 0610-03
. D E FI NITIO N S................................................. 1 i
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0610-04 R E S PON SI BI LITI E S............................................ 3
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04.01 General Responsibilities-Reactor,.................................. 3 j
04.02 Report W riting............................ '.................... 3 04.03 Report Review and Concurrence................................... 3 04.04 Report lasuance................................................ 4 04.05 Report Timeliness.............................................. 4 0610-05 GUIDANCE - INSPECTION REPORT CONTENT...................... 4 05.01 Observations, and Findings,...................................... 4 a.
Observations.................................................. 5 b.
Findin gs...................................................... 5 05.02 Thresholds of Significance........................................ 7 a.
Thresholds of Significance for Noncompliance issues............................../......................... 7 b.
Threshelds of Significance for Non-Enforcement-Related i ssues................................................. 7 05.03 Level of Detail................................................. 8 l
a.
W ho is the Reader?............................................. 8 b.
Importance of Overall Conciseness................................. 9 c.
Level of Detail on inspection Scope................................. 9 d.
Level of Detail on Observations and Findings........................ 10
~ 05.04 Documenting Noncompliances.......,............................ 10 i
a.
Specific Enforcement Related Guidance............................ 11 lasue Date: DRAFT 5/19/99 -
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r b.
Supporting Dstrils and Discussions of Siftty Sig nifica nce.................................................. 12 c.
Noncompliances involving Willful 1ess.............................. 13 05.05 Documentation of Performance-E,ased inspection..................... 13 a.
Documenting Performance-Basr S istues vs.
Compliance-Based Issues....................................... 13 b.
Documenting issues in Areas Not Covered by Regulatory Requirements....................................... 14 c.
Documenting Weaknesses (Exercise Weaknesses)................... 17 d.
Documenting Management issues................................. 17 05.06 Treatment of Open items in Reactor Inspection Reports................ 18 a.
Initiating Open items........................................... 18 b.
Follow-Up and Closure of Open items.............................. 18 c.
Treatment of Licensee Event Reports.............................. 18 d.
Avoiding " Implied" Inspection Follow-Up Items....................... 19 0610-06 GUIDANCE - INSPECTION REPORT FORMAT...................... 19 06.01 Cove r Lette r................................................. 19 06.02 Cover Pa g e.................................................. 20 06.03 Summary of Findings / Plant Information Matrix........................ 20 06.04 Table of Contents............................................. 23 06.05 Report Details: Use of the Standardized Report Outline................ 23 06.06 Report Details: Intemai Organization of Specific Sections..................................................... 23 06.07 Exit Meeting Summary.......................................... 24 06.08 Report Attachments............................................ 25 0610-07 GUIDANCE - INSPECTION REPORT STYLE........................ 25 07.01 Peculiarities of Government Technical Writing........................ 26 07.02 Clear Organization............................................. 28 07.03 Effective Revision............................................. 29 0610* DRAFT
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07.04 Writing Styl 3 Guides tnd Useful NRC R;far:nc:s..................... 31 0610-08 RELEASE AND DISCLOSURE OF INSPECTION REPORTS AND ASSOCI ATED DOCUMENTS.................................... 31 Exhibit 1: Standard Reactor inspection Report Outline........................... E1-1 Exhibit 2: Sample Reactor Inspection Report................................... E2-1 Appendix A: Noncompliance Information Checklist............................... A-1 Appendix B: Managing the Writing Process..................................... B-1 i
Appendix C: Inspection Report Review Checklist................................. C-1 Appendix D: List of Acronyms Used in this inspection Manual Chapter................ D-1 l
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INSPECTION REPORTS 0610-01
-PURPOSE
- To provide guidance on reactor baseline inspection report content, format, and style.
0610-02 OBJECTIVES To ensure that inspection reports:
02.01 Clearly communicate significant inspection resulu to licensees, NRC staff, and the public.
02.02 Provide a basis for enforcement action.
NOTE: NUREG-1600 is the NRC Enforcement Poliev. Enforcement guidance is discussed in -
Section 05.04 02.03 Assess licensee performance in a periodic, short-term context, and present information in a manner that will be useful to NRC management in developing longer-term, broad assessments of licensee performance.
0610-03 DEFINITIONS Apparent violation. A potential noncompliance with a regulatory requirement that has not yet been formally cited as a violation in a Notice of Violation or order.
Closed item. A matter previously reporicd as a noncompliance, an exercise weakness, a licensee event report, or an unresolved item, that the inspector concludes has been satisfactorily resolved, based on information obtained during the current inspection.
Deviation. A licensee's failure to satisfy a regulatory commitment.
Draft inspection Report. Any version of the inspection report before its official issuance.
1 Escalated Enforcement Action. A Notice of Violation for any Severity Level I, ll, or til violation (or problem), or a civil penalty or order based on a violation.
Exercise Weakness (EW). A finding that the licensee's demonstrated level of preparedness could have precluded effective implementation of the emergency plan in the event of an actual emergency (see 10 CFR Part 50, Appendix E, Section IV.F.2.g).
i Findina. ~ As used in this chapter, an observation that has been placed in context.
Inspection.
The examination and assessment of any licensee activity to determine its ofOctiveness, to ensure safety, and/or to determine compliance. A single inspection report may e compass resident inspection, in-office document review, and/or one or more visits by regional or headquarters inspectors; however, a single report is normally limited to a specific period of inspection (e.g., a 6-week period),
inspection Document. Any material obtained or developed during an inspection that is considered to be an NRC record (see below).
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Insoection Follow-Up Item. [lFis will not be u ed in ths Bas lins insp:ction Progr:m)
IntearatedInspectionReoorts. Areactorinspectionreportthatcombinesinputsfromallinspections (resident, regional, etc.) conducted within a specific period.
Licensee. The applicant for or the holder of an NRC license, construction permit, or combined license.
Minor V,olation.
A violation that is less significant than a Severity Level IV violation (previously categorized at Severity Level V), not the subject of formal enforcement action, and not usually described in inspection reports or inspection records.
Noncited Violation (NCV). A violation for which the staff chooses to exercise discretion and refrain from issuing a 10 CFR 2.201 Notice of Violation.
Noncomoliance. A violation, noncited violation, deviation, or nonconformance.
Notice of Violation (NOV). A formal written citation in accordance with 10 CFR 2.201 that sets forth one or more violations of a legally binding regulatory requirement.
NRC Record. Any written, electronic, or photographic record under legal NRC control that documents the policy or activities of the NRC or an NRC licensee (see also the definition in 10 CFR Part 9).
Observation. A fact; any detail noted during an inspection.
Open item. A matter that requires further inspection. The reason for requiring further inspection may be that the matter has been identified as a noncompliance, unresolved item' or licensee event report.
Potentially Ger,eric issue. An inspection finding that may have implications for other licensees, certificate holders, and vendors whose facilities or activities are of the same or similar manufacture or style.
Reculatory Commitr,;ent. An explicit statement to take a specific action, agreed to or volunteered by a licensee, where the statement has been submitted in writing on the docket to the NRC.
Reauiremen_t. A legally binding obligation such as a statute, regulation, license condition, technical specification, or order.
Sionificance Determination Process (SDP). The process used to determine the risk significance of inspection findings.
Unresolved item. A matter about which more information is required to determine whether the issue in question is an acceptable item, a deviation, or a violation.
Vendor. A supplier of products or services to be used in an NRC-licensed facility or activity. In some cases, the vendor may be an NRC or Agreement State licensee (e.g., nuclear fuel fabricator, radioactive waste broker) or the vendor's product may be required to have an NRC Certificate of Compliance (e.g., certain transport packages such as waste casks cr radiography devices).
Violation. The failure to comply with a legally binding regulatory requirement, such as a statute, regulation, order, license condition, or technical specification.
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Willfulne_s_q. An cttitude tow rd compliines with requiram:nts that rang:s from th3 car:l:ss disregard for requirements to a deliberate intent to violate or to falsify.
0610-04 RESPONSIBILITIES All NRC inspectors are required to prepare inspection reports in accordance with the guidance provided in this Irspection Manual chapter. General and specific responsibilities are listed below.
04.01 General Responsibilities--Reactor. Inspections: Each inspection of a reactor should be documented in a report consisting of a cover letter, a cover page, an summary of findings, and inspection details.
04.02 Reoort Writina a.
Inspectors have the primary responsibility for ensuring that observations and findings are
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accurately reported, and that referenced materialis correctly characterized. Advice and j
recommendations are not to be included in inspection reports.
b.
Inspectors are responsible for ensuring that the content and tone of the report, as issued, is consistent with the content and tone of the exit meeting presentation. When the report differs significantly from the exit meeting, the inspector (or the report reviewer) should discuss those differences with the licensee before the report is issued.
c.
Report writers and reviewers should en:,urs that inspection reports follow the general format given in this chapter and in the enclosed sample report (see Exhibits 1-2).
d.
For inspections conducted by regional and resident inspectors, the report number is in the following form:
Docket No/ Year - sequential number of the report in that year (e.g.,50-363/99-01)
For inspections conducted by NRR, NMSS, etc., the report number is in the following form:
Docket No./ Year - 2 followed by the sequential number of the report in that year (e.g.,
50-250/99-201)
NOTE: The report number format given here is for use in the inspection report itself.
~ This format may be modified as needed for other applications (e.g., for item Report and Analysis Module (IRAM) entries), because of electronic constraints and, other considerations.
04.03 Report Review and Concurrence a.
Befora issuance, each inspection report should, as a minimum, be reviewed by a member of NRC management familiar with NRC requirements in the area inspected, b.
The report reviewer (i.e., the member of management referred to above) should establish that the observations and findings are consistent with NRC policies and requirements.
c.
The report reviewer should ensure that assessments made in the inspection report are in accordance with the SDP.
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d.
R:gion:1 cdmini;trators and offica dir:ctors should establish int:rnal proc:dur:s to provide a record of inspectors' and reviewers' concurrences. The procedures chould 4
address how to ensure continued inspector concurrence when substantive changes are made to the report as originally submitted, and how to treat disagreements that occur during the review process. As a minimum, substantial changes should be discussed with the inspector or inspectors involved to ensure continued concurrence, and disagreements that cannot be adequately resolved shou':1 be documented.
NOTE: The record of inspector and reviewer concurrence is maintained by the issuing office. This concurrence record is not included in the generally distributed version of the report.
04.04 Report issuance a.
For regionalinspection reports, the applicable division director or designated branch chief is responsible for the report content, tone, and overall regulatory focus. For integrated reports issued to reactor licensees, the Division of Reactor Projects (DRP) division director or designated branch chief is responsible for issuing the report to the licensee.
04.05 Report Timeliness a.
General Timeliness Guidance, inspection reports should be issued no later than 30 calendar days after inspection completion. (45 calendar days for integrated reports and major team inspections).
NOTE: Inspection completion is normally defined as the day of the exit meeting. For resident inspector and integrated inspection reports, inspection completion is normally defined as the last day covered by the inspection report.
b.
Reports Precedina Escalated Enforcement Actions.
Timeliness goals should be accelerated for inspection reports covering potential escalated enforcement actions. For specific enforcement timeliness goals, see the NRC Enforcement Manual, c.
Expedited Reports for Sianificant Safety Issues. Whenever an inspector identifies an issue involving significant or immediate public health and safety concerns, the first priority is facility and public safety; issues of documentation or enforcement action are secondary.
Based on the circumstances of the case, an expedited inspection report may be prepared that is limited in scope to the issue, or expedited enforcement action may be taken before the inspection report is issued. The NRC Enforcement Manual provides additional guidance on matters of immediate public health and safety concern.
0610-05 GUIDANCE INSPECTION REPORT CONTENT This section relates primarily to matters of content in the inspection report details. Some guidance on the content of report cover letters and Summarv of Findinos is given in Sections 06.01 and 06.03, respectively.
05.01 Observations and Findinos. As used in this chapter, tne term " observation" refers to a fact--
any detail noted during an inspection. The term " finding" designates an observation that has been placed in context.
Adherence to the use of these terms is less important than appreciating the underlying process.
Achieving relative consistency in inspection report content first requires an understanding of how 0610* DRAFT Issue Date: DRAFT 5/19/99
inspection obs:rv:tions cra cssess:d for significanc3 by th3 SDP, cnd how th3 resulting findings should be docum:nted.
a.
Observations. The most basic results of an inspection are the facts an inspector gathers--
through watching work activities, examining equipment, interviewing licensee employees, reviewing records, and other inspection methods. As documented, these observations should be factual--that is, an inspector should not report hunches, unsubstantiated hearsay, or unverified opinions.
Consider an inspection in which a maintenance worker states that a particular pump has excessive shaft leakage, and the inspector wants to make this an inspection observation.
If possible, the inspector should physically observe the pump to verify the maintenance worker's statement. If the pump is inaccessible, or the leakage has already been corrected, the inspector should attempt to verify the period of excessive leakage through log review, maintenance records, discussions with additional plant personnel, or some other method.
When documenting an observation, use language that clearly identifies how the observation was discovered and verified (see also the discussion on active voice in Section 07.01.b.3):
BAD: "The chemical and volume control system (CVCS) make-up pump had excessive shaft leakage," OR "It was noted that the CVCS make-up pump had excessive shaft leakage."
GOOD: "A maintenance worker informed the inspector that the CVCS make'-up pump had excessive shaft leakage. The inspector verified this information by observing the pump in operation," OR "... by reading the operator logs for June 5-6."
NOTE: Neither of these statements explain the significance of the observation, nor the criterion used.
Other factuat information may be relevant to an observation, such as the time of discovery,'
the length of time the problem existed, the type, size, or model of the equipment and whether the pump is unavailable, and so forth. If the pump is unavailable, obtain the necessary information to make an SDP analysis, for example; availability of attemate equipment and any changes in risk due to the unavailability. Section 05.03 discusses how to determine the appropriate level of detail.
b.
Findinas. Although inspectors are frequently told to " bring back the facts," NRC managers also expect those facts to be placed in context, so that the SDP can be used for assessment. Reports should be free of " dangling observations"-that is, items that leave 1
the reader unclear as to (1) how the observation relates to a requirement or standard, or (2) what factors were used in arriving at the SDP determination. By answering these i
questions, inspection observations become inspection findings.
j 1.
Referencina Reauirements and Standards. Whenever possible, an observation should be related to a requirement or standard. Often this context is achieved by direct numerical companson:
EXAMPLE: "The inspector observed, on control room Panel R442, that service water flow through residual heat removal (RHR) Heat Exchanger B was indicating approximately 5900 gallons per minute (gpm). The surveillance rninimum service water flow, as given in TS 4.5.2.e, is 6500 gpm."
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For soma observitions, the stInd rd'to be rafar:nced will be aualitative. Th se observations sb^uld normally be limited to observations (or findings) which are SDP candidates. For example, a finding that the licensee unknownly made a measurable increase in plant risk because of a combination of work items, for which no TS violation occurred but a operating reactor SDP phase 2 screening was required, should be discussed in a report.
2.
Clanfying Findings. The reader should be left with a clear sense of what the finding is. Often this is made obvious by the context, or by relating the observation to a requirement or standard. If the observation is " neutral" and does agt relate to a requirement or standard, the inspector should question whether the item is significant enough to be documented at all. Neutral observations should be very limited. In some cases, a significant " neutral" observation can be placed in context by its relation to past performance:
EXAMPLE: " Total extemai exposure for Refueling Outage 2R6 was 321 rem, which slightly exceeded the outage goal of 315 rem. The total extemal exposure received was 15 percent higher than the total for the previous outage. According to the licensee's analysis, this difference was primarily due to the extensive steam generator tube sleeving and plugging performed during Outage 2R6."
3.
Providina Context for Sionificance.
Fully assessing the significance of an observation may require consideration of many factors as directed in the SDP. In addition, on a case-by-case basis additional information may help with proper characterization of the finding: Who was involved in the issue? Has this occurred before? Is a trend or pattem developing? Who found the problem? Did the licensee have an opportunity to discover an issue sooner? Has the licensee entered.the issue in their corrective action system? How does the licensee characterize the significance of this matter?
The report need not always answer each of these questions, and need not exhaustively provide every supporting detail for every observation. The inspector should weigh the circumstances impartially, and should include in the report those details that contribute to understanding the significance of the observation--
regardless of whether they make the finding appear more severe or more benign.
In all cases, the report should address the factors required for proper SDP determination. However, for observations that are immediately determined to be i
within the licensee's response band (green) during initial SDP screening, less detail i
is required.
NOTE: An inspector should always document supporting details for findings. On the other hand, inspectors should be careful not to make direct statements, in the report details, regarding the safety significance of the noncompliance which maynot be consistent with the SDP determination.
Since the process of assessing significance (or translating observations into findings) can be subjective, it requires skill, experience, and judgment, and demands that the inspector carefully consider all viewpoints. The inspector should make every effort to understand and fairly characterize the licensee's perspective.
In addition, the inspector's final assessment of a finding's significance, as determined by the SDP, should be apparent to a knowledgeable reader.
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05.02 Thresholds of Significance. At most lic:nsed ficiliti:s, ev ry day of operation involv2s hundreds of peopla, doz:ns of svolutions, and multipla pitnt syst:ms. Probl:m3 cra id:ntifi:d end resolved daily; most licensees process large numbers of " Corrective Action Requests" or similar problem tracking mechanisms. Within this scenario, the NRC inspector essentially performs an auditing role. The inspector cannot hope to monitor all the activities in progress, nor to document every minor discrepancy that occurs. As part of maintaining a focus on safety, inspectors continually use NRC requirements, inspection procedures, risk assessment documents, industry standards, regional and headquarters guidance, and their own training and insight to make judgments about which issues are worth pursuing and which are not.
To communicate effectively,-inspection reports must give evidence of that judgment and prioritization, discussing significant safety issues in appropriate detail, treating less significant issues succinctly, and avoiding excess verbiage. To maintain some consistency in how lesues are treated, report writers must recognize certain " thresholds of significance": that is, they must use similar criteria in deciding whether an issue is important enough to document a.
Thresholds of Sionificance for Noncomoliance issues. See Section 05.04 for guidance on enforcement for the baseline inspections.
Minor Violations-Determinina Whether to Document. In general, minor violations should Dgi be documented; however, certain exceptions apply. Documentation may be necessary as part of the resolution to an allegation. In other cases, while the violation itself is minor,
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the associatea technical information may relate directly to an issue of agency-wide concem (e.g., the inspection was performed in response to an NRC Temporary Instruction (TI)). If, for these reasons or any other reason, the report writers and reviewers wish to document a minor violation, then it should be documented as a minor violation, with a reference to Section IV of the NRC Enforcement Poliev. For example, "This failure constitutes a violation of minor significance and is not subject to formal enforcement action." Minor violations shall not be included in the Summary of Findings and shall not be given a new number. If an issue already has an EA number or other tracking number, and is determined to be minor, it is acceptable to use the existing number when discussing the minor violation.
b.
Thresholds of Signifmance for Nonenforcement-Related issues.
Inspectors must also make judgments about the relative significance of nonenforcement-related findings. As with enforcement issues, the judgment of individual inspectors will differ; questions on the relative significance of an issue should be discussed with other inspectors and with NRC managers.
1.
Determinina the Sionificance of Neoative Findinos. The following questions should be used to determine whether or not a finding should be documented in the inspection report:
1 Does this finding have any actual impact (or any significant potential for impact) on safety?
Is this finding illustrative of a problem that could have a safety or regulatory impact?
j Does this finding provide insights on an equipment, system, or human performance i
problem?
i Could this finding be viewed as the possible precursor to a more significant event?
If the licensee takes no action on this matter, will the condition worsen (i.e., will the i
safety significance increase)?
4 If this finding recurs, will its recurrence result in more significant or additional safety concems?
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Will thl2 inform: tion be us:ful in ess:ssing th] long-t:rm p;rform:nc3 of th3 o
licensee's corrective action program? Because the baseline inspection program relies on an adequate licensee corrective action program, program related findings associated with identification and resolution of problems should be included in the report when the answer to this question is a "yes." For other licensee programs, findings will not be included unless they address the other questions in this section.
Does this finding have generic significance?
If the answer to any one of these questions is "yes," the finding may be documented in the inspection report. If the answer to all questions is "no," the finding normally should not be documented.
2.
Use of Neutral Observations. Neutral observations will not normally be used. If neutral observations are used, higher thresholds of significance should apply. The inspector should ask questions similar to those below:
Does this information provide useful equipment, system, or human performance j
information?
Does this licensee action significantly reduce the probability of a particular event?
Will this information be useful in assessing the long-term performance?
Does this observation have generic significance?
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If the answer to any one of these questions is "yes," the observation may be documented in the inspection report. If the answer to all questions is "no," the observation normally should not be documented. Neutral observations should not be included in the Summarv of Findinas, except for the corrective action program observations required by the yearly corrective action inspection.
3.
Findinas Previousiv Covered in Licensee Self-Assessments.
In general, little benefit exists in NRC's re-emphasis of issues already covered in licensee self-assessments, unless there is some problem with the licensee's actions.
In some instances, however, the technical significance or generic implications of an issue merit ensuring that it is discussed on the docket and preserved as a matter of public record. If the licensee self-assessment that initially discussed the issue is already on the docket, the inspection report may simply refer to the discussion in the licensee self-assessment. If more detailis needed, orif the licensee self-assessment is not on the docket, the inspector may wish to discuss the issue in the inspection report narrative. in general, discussion of these type of issues in inspection reports should be limited to SDP candidates, significant issues not covered by any SDP, or yearly corrective action program inspection results.
05.03 Level of Detail. Just as inspectors must use judgment in determining what issues are worth including in the inspection report, so they must also determine the appropriate level of detail for issues that.a.Le included. Some issues should be discussed in more detail than others, based on safety or regulatory significance, technical complexity, and other factors.
a.
Who is the Reader? In writing any technical or business document, the most basic step in ensuring the appropriate level of detailis to have a clear sense of the reader--his or her background, priorities, and level of expertise in the subject area. Inspection reports have multiplereaders,withvaryinglevelsof technicalexpertise(includingindividualsatalllevels of the NRC and licensee staffs, other licensees, vendors, industry groups, public interest groups, and members of the general public). The principal reader, however, is the person to whom the report cover letter is addressed.
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For r=ctor licensns, this is g ;nt rally th a vic2 pr:sid:nt, nucl=r, or a similar high-ranking comprny officirl. Th3 r: port should be written, therefore, with a corresponding level of technical detail, so that it will be understood by a knowledoeable individual conversant with nuclear technoloov. but who may not be an expert in the specific area inspected.
Plant-specific design features, relevant procedures, event-specific details, and other factual information should be presented in sufficient detail to allow this " target audience" to understand the characteristics and significance of the inspection findings.
NOTE: See also the discussion on using technical, legal, and local jargon in Section 07.01.d.
Finally, the report writer should consider that he or she is writing for the " record"--the docket file that chronicles official NRC interactions with the licensee. While the " record" is not a reader in the usual sense, the writer should be aware that later readers who use the docket file for research and historical perspective will frequently be dependent on the level of detail in an inspection report. Certain types of detail can be especially helpful for these readers: knowing the inspection procedure used, the exact component (or system or train) inspected, the component manufacturer (where relevant), the revision number of a referenced licensee procedure, the date, time, and duration of a plant event or transient, and similar details. This awareness--that one is writing for the record-should not be taken as an incentive to write long narrative descriptions, but it should provide motivation for writing precisely.
b.
Imoortance of Overall Conciseness. For most writers, the second step in ensuring an appropriate level of detail lies in learning to differentiate between information that contributes to understanding the findings and information that detracts or merely adds verbiage. The details used to describe any given finding should be scrutinized from this standpoint, and the nonessential details pared away. If properly performed, this process will result in a clearer presentation of the findings and a better communicated message overall.
In leaming to write concisely, many inspectors will need to overcome a widely held but naive assumption that a shorter (or less thick) inspection report is a report of less merit.
The report is not intended to be a lengthy discourse of activities carried out so as to justify J
the time spent or to demonstrate knowledge of a particular technical area. For any given inspection, if no safety or noncompliance issues were identified, then there is no need to document this inspection except as discussed in Section 06.06.
c.
Level of Detail on Insoection Scope. The level of detail here should be minimal and specifically focus on what was actually inspected. Inspectors should not fallinto the habit of inserting long "boilerplate" paragraphs at this point to establish that the inspection program is being implemented:
INAPPROPRIATE INSPECTION SCOPE EXAMPLE: "The inspectors performed walkdowns of the accessible portions of various engineered safety feature (ESF) systems. In performing these walkdowns, the inspectors verified the proper installation of hangers and supports; the adequacy of housekeeping; correct valve position and conditions; the absence of ignition sources; proper labeling, lubrication, and cooling of major components; operational status of support systems, including instrumentation; consistency of significant process parameter values with expected values..."
issue Date: DRAFT 5/19/99 0610* DRAFT
s 2 Descriptions of this sort serva littia purposa. Th:y crs u:uilly copied from previous inspection reports or paraphrased f rom the inspection procedure, and they are rarely read.
Detailed descriptions of inspection methods or of "what was inspected" should only be
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included when they'are needed to understand or add perspective to the inspection findings. A tabular format is frequently useful:
APPROPRIATE INSPECTION SCOPE EXAMPLE: The inspectors checked equipment alignments for the accessible portions of the following ESF systems:
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Emergency Diesel Generator 2 3
High Pressure Coolant injection j
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This level of detail is normally appropriate when reporting a review of licensee procedures, i
observed work in progress, routine plant evolutions monitored, or similar inspection i
activities.
For certain types of inspection activities, more detail is appropriate, for example; When the inspector is present during a significant plant event or. an unusual plant evolution, more detail may be appropriate conceming which portions of the event or evolution were actually observed.
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d.
' Level of Detail on Observations and Findinos. Once the inspector has decided that a
' finding is important enough to be included in the report, the same questions used in
' making that decision (see Sections 05.01.b.2 and 05.02) can assist in determining the appropriate level of detall. The following guidance applies.
.1.
The degree of actual or potential safety consequence associated with a finding I
'should be a primary consideration in determining the level of appropriate detail.
Items of higher significance (reactor SDP phase 2 screening or similar issues) generally merit more discussion.
2.
Findings of greater technical significance--that is, findings that give insights into equipment, system, or human performance issues, or findings that could have generic significance-should be discussed in sufficient detail to communicate those
'nsights.
3.
When the inspector has identified that a particular finding has added significance based on risk, that perspective should be explained. For example, if the inspector finds that two components with reliability problems are related by a dominant event sequence, that relationship should be explained.
4.
As a general rule, " neutral" assessments should be described in less detall than negative findings. Additional " neutral" details may be warranted when reporting certain performance indicators or similar information that will be useful in assessing long-term performance. Neutral assessments on the adequacy of the licensee's corrective action program should be included during the annualinspection of this area.
- 5.
Positive findings should not be documented. However, when describing all the information that was needed to properly perform an SDP, findings that licensee 0610* DRAFT Issue Date: DRAFT 5/19/99
cctions w:rs etdequits to mitig ta a probl:m should be rupport:d by tha b: sis for i
that finding.
6.
When initiating an unresolved item the issue description should provide enough background information that a different inspector, using that information, would be equipped to perform the follow-up inspection.
)
05.04 Documentina Noncomoliances.
The primary guidance for all matters related to
)
enforcement, including documentation, is given in the NRC Enforcement Policy (NUREG-1600),
and the blRC Enforcement Manual (NUREG/BR-0195).
This guidance applies to issues found or reviewed during inspections that are also violations of requirements. The focus of the activities delineated below is on determining the significance of the issues and assuring that the licensee has taken actions appropriate for the issue. The significance determination process (SDP) will be used, where applicable, for making the determination of significance. Issues that are not evaluated under the significance determination process will be processed in accordance with the traditional enforcement process. Such issues are situations with actual safety consequences (such as an overexposure to the public or plant personnel or a substantial release of radioactive material) or are violations related to willfulness or to impeding the regulatory process. (Ref EGM/ Enforcement Policy) a.
Soecific Enforcement Related Guidance
-1.
For issues that are determined to be within the licensee response band (i.e., green),
the associated violation will be treated as noncited in accordance with the Enforcement Policy. The noncited violation will be documented in the associated inspection report and entered into the Summary of Findinas and plant issues matrix.
The discussion of NCVs should include sufficient information to support the violation was not minor.
2.
For issues that are determined to be significant (i.e., white, yellow, red), the following steps will be taken:
The assumptions used by the inspector or regional SRA in determining the issue's significance will be documented.
The' licensee will be informed of the issue, its significance level, and the assumptions used by the NRC in determining significance.
The licensee is asked to provide three pieces of information:
The significance attributed to the issue by the licensee and, if different than the NRC's significance level, a description of the assumptions the licensee used and considers applicable to its determination.
Actions the licensee has taken or plans to take to correct the condition and underlying root cause(s).
The licensee's position on the NRC's determination that a requirement has been violated.
The NRC reviews the licensee's response and conducts a phase 3 significance determination process review, if necessary, to reach a final position on the issue's significance and the appropriateness of the corrective actions.
Issue Date: DRAFT 5/19/99 0610* DRAFT
j Th3 finil significance d:t:rmin: tion is docum:nted, ths issu3 is ent r:d into i
the plant issues matrix, and the associated enforcement action is taken based l
on the significance:
If the issue is green, a noncited violation is documented in an inspection report.
l If the issue is white, yellow, or red, a notice of violation is issued in l
accordance with the Enforcement Policy /EGM.
3.
The preceding steps should be timely and whenever possible accomplished during i
the inspection period so the final determination and associated enforcement action l
can be issued with the associated inspection report. In those instances when that 1:., not possible (such as when the issue is found late in the inspection period or determining its significance needs specialized help), the following will be done:
The issue will be discussed in the inspection report and identified as an
" apparent violation."
l The basis for the NRC's significance determination and the request for l
information (Section 05.04.a.2 above) will be sent to the licensee by letter as soon as the determination has been made. The licensee should be given at least 2 weeks to respond. Any meetings with the licensee needed to discuss the issue or the licensee's response must be open to the public.
The next regularly scheduled inspection report will document the disposition of the issue if the final determination results in a noncited violation or a finding that no requirements were violated.
A separate letter with a notice of violation will be sent to the licensee if the issue is finally determined to be of greater significance. In general, the licensee will be requested to respond to the notice of violation unless the information received from the earlier request adequately addresses corrective actions for the issue. Additional meetings may be requested by either the licensee or the NRC if warranted by the specific issue.
i 1
b.
Supportina Details and Discussions of Safety Sianificance. The discussion of findings must be sufficiently detailed to substantiate any NRC safety and regulatory findings and i
to support any enforcement sanction the NRC may choose to issue. The narrative should answer, as appropriate, the questions in the SDP instructions. Additional questions for consideration are given in the Noncompliance Information Checklist (see Appendix A).
At a minimum, for a violation or NCV, the report should state:
4
- what requirement was violated;
- how the violation occurred;
- when the violation occurred, and how long it existed;
- when the violation was identified;
- any actual or potential safety consequence;
- the root cause (if identified);
- allinformation required to complete the SDP;
- what corrective actions have been taken or planned. (For licensee's with adequate corrective action programs, it is acceptable to only verify that the licensee has entered i
the issue in its corrective action program).
0610* DRAFT Issue Date: DRAFT 5/19/99
Th3 degram of detiil n:crsary to support en enforc:m:nt cction is a function of tha significance and complexity of the noncompliance.
Although supporting details clearly assist in determining the safety significance of the noncompliance, inspectors must be careful to avoid making direct statements regarding safety significance in the inspection report details outside the SDP analysis. Violation severity levels, as described in the NRC Enforcement Poliev. are based on the degree of safety significance invcIved. In addition, the NRC EnSrcement Policy uses the term
" safety significance" in a specific sense, which involves consideration of (1) actual safety consequence, (2) potential safety consequence, and (3) regulatory significance (e.g.,
willfulness or management involvement in a noncompliance, programmatic breakdowns, repetitive violations, etc.). Inspection reports should not refer to a noncompliance as being "of low safety significance" (meaning, in a general sense, that the noncompliance did not result in any actual adverse impact on plant equipment or personnel), because the SDP will, in effect, determine the significance of all issues.
c.
Noncomoliances involvina Willfu' ness. Inspection reports should neither speculate nor reach conclusions about the intent behind a violation, such as whether it was deliberate, willful, or due to careless disregard. As with any observation, the report discussion should j
include relevant details on the circumstances of the violation without making a conclusion about the intent of the violator:
APPROPRIATE: "The radiographer failed to activate his alarming dosimeter, although he had informed the inspectors earlier that he had been properly trained on the use of the device."
INAPPROPRIATE: "Ths radiographer deliberately failed to activate his alarming dosimeter."
Conclusions about the willfulness of a violation are agency decisions, and are normally not made until after the Office of Investigation (01) has completed an investigation and a predecisional enforcement conference has been held. A premature or inaccurate discussion of the willfulness of an apparent violation in the inspection report could result in later conflicts based on additional input and review. Inspection reports that include potentially willful violations are to be coordinated with Ol and the Office of Enforcement (OE).
05.05 Documentation of Performance-Based Insoection. " Performance-based inspection" is inspection that focuses on issues of safety and reliability, with an emphasis on field observation rather than in-office procedural or record reviews. Tne emphasis on safety and reliability f requently borrows from risk studies, incorporating probabilistic risk assessment (PRA) and individual plant i
examination (IPE) insights to structure inspections that focus on systems or components most I
important to plant safety. In addition, performance-based inspectiori tends to focus more on results (e.g., does the pump work?) than on process and method (e.g., was the pump maintenance prccedure well-written?).
In moving toward " performance-based regulation," more recently developed NRC requirements (such as the maintenance rule,10 CFR 50.65) tend to be less prescriptive about process or method and more focused on results than earlier regulations. For most areas of inspection, the range of relevant regulations, license requirements, industry guidelines, and licensee regulatory commitments is a mixture of performance-based (results-oriented; less prescriptive) and compliance-based (process-oriented; more prescriptive) standards. This mixture often makes it difficult for inspectors / report writers to present and document inspection findings in a consistent manner.
Issue Date: DRAFT 5/19/99 0610* DRAFT I
l
1 2.
Documentaa Performance-Based issues vs. Comoliance-Based issues. Th3 first st:p in documenting ' performance-based' findings is understanding the underlying flow of logic, and differentiating this logic from that of a finding based strictly on compliance. For compliance issues, the clearest manner of presentation is usually comparison / contrast, similar to the format of an NOV.
EXAMPLE: Suppose the inspector finds, that a certain surveillance is not being
' conducted at the required frequency. No performance problems exist with the equipment, and licensee follow-up of the observation determines it to be an isolated area of operator oversight, with no underlying training or procedural problem. The inspector might present such a finding in the following manner:
"TS states that the instrumcnt channel shall be verified operable by performing CHANNEL CHECK and CALIBRATION operations at frequency. However, from April 7,20_ until the inspector identified the issue on August 13,20_, the CHANNEL CHECK and CALIBRATION operations were only performed at a frequency of
. thus failing to meet the above requirement..." followed by a brief summary of the inspector's follow-up actions (if any), the licensee's response, the SDP results, and concluding with statements that disposition the viole. tion.
By contrast, a performance-based finding frequently begins with the field observation of a safety or reliability issue (e.g., an equipment problem, a deficient work practice, a questionable system response, etc.), which results in efforts to place the observation in context, understand any associated problems with the underlying processes or methods--
all of which may or may not lead to an issue of noncompliance. When documenting such a finding, the clearest presentation usually follows the same path of discovery-that is, the j
narrative (1) begins with a statement of the observation, (2) places that observation in the i
context of related findings or circumstances that contribute to understanding its j
significance, (3) explains any known root causes or underlying process problems, (4) leads to a " bottom-line" finding that a particular standard was or was not met (if the standard is a requirement, this may be a finding of noncompliance) and (5) provides the necessary information for an SDP determination.
This performance-based approach can be a factor in determining whether an observation is important enough to documeat, and if so, what level of detail is appropriate. For '
example, the organization and staffing of a particular licensee group is seldom an appropriate topic from which to build significant findings. Few NRC requirements relate to organization and staffing; as a stand-alone issue, it rarely merits a dt.M ed report il discussion.
b.
Documentina Issues in Areas Not Covered by Reculatorv Reauirements. Although the NRC always seeks to focus the requirements of its regulations and licenses on safety consicarations, mere compliance with those requirements does not automatically ensure safety. The NRC's safety mandate entails inspection and evaluation of licensee performance in areas that may not be covered by written requirements. In general, only Issues which are initially considered for SDP review (SDP candidates), or equivalent issue-s in areas not currently addresses by SDPs, should be included in the report.
Presumably, judgments made in this realm-in areas not covered by NRC requirements--
must still use some standard as a reference point. Various inspection procedures give specific criteria for the inspector to use in evaluating a licensee's performance-including some criteria that are not directly related to an NRC requirement, and that might be more correctly characterized as matters of industry convention or standard nuclear safety practices. When inspection findings are made in these areas-that is, when safety issues 0610* DRAFT Issue Date: DRAFT 5/19/99
tra id:ntified thit do not rtista directly to a regulatory requirement--ths treatm:nt of such findings can be extrsmaly difficult. How ara such findings to be dispositionsd at an exit meeting, or in an inspection report?
1.
Avoidina Makina Recommendations or Creatina New Reauirements. As the first
" rule of thumb" in this area, note that the " standards" discussed here are aeneraljy recoanized princioles of safe ooeration. and are not written or stated in a manner to resemble concrete reauirements. For example, the generally recognized principle of keeping exposures as low as is reasonably achievable (ALARA) justifies writing the following statement, ACCEPTABLE: " Licensee conduct of work in radiologically controlled areas should give evidence of in-process controls to minimize radiation exposure."
It would Dgi be appropriate to prescribe specific in-process controls that constitute recommendations or could be construed to be new requirements:
UNACCEPTABLE: " Licensee conduct of work in radiologically controlled areas should include remote monitoring cameras and/or direct job-site supervision by a radiation protection technician."
Since a focus of the baseline inspection program is to ensure that the licensees are properly managing risk, it is expected that there will be findings where configuration control or similar errors lead to increased risk. Even if no direct NRC requirement exists, these type of findings should be considered for inclusion in inspection reports, based on the threshold of significance of the finding.
1 Note also that, when seeking to establish a clear standard of expected performance in areas not covered by NRC requirements, inspectors must be careful never to imoose oersonal oreferences or arbitrary ooinions on the licensee. Standards of expected performance should be discussed with both NRC and licensee management, and the inspector should promptly bring any licensee disagreements to the attention of NRC management.
NOTE: See also the discussion of backfits under Section 05.05.b.4.
2.
Usina Standards in Areas not Covered by NRC Reauirements. The inspector should attempt, through review of inspection procedures and discussions with NRC and licensee management, to arrive at a clear statement of expected performance.
That statement should then be included in the report narrative.
EXAMPLE: Suppose that the licensee identifies the failure of the "A" containment fan cooler motor inside the containment, and decides to troubleshoot the motor during an "at power" entry. The inspector monitors this maintenance activity, and makes the following observations:
While waiting to make the containment entry, the inspector notes that the prejob briefing placed little emphasis on the actual work to be performed or the caution statements included in the work package; At the job site, the inspector notes that initial communications with the control room were confusing and hard to hear due to in-plant noise; As the maintenance workers were about to begin dismantling the motor, the inspector observes that they were going to work on the "B" motor rather than the failed "A" motor, and alerts the workers to this problem; lasue Date: DRAFT 5/10/99 0610* DRAFT l
)
Becaus3 of thm d: lays, errors, cnd th3 resulting cdditional tima and effort, ths r:diation exposurs r:crivid w:s niarly 75 mr m mora th n plann:d.
Each of these observations are valid and insightful, yet the inspector may in each case be unable to establish that any requirement has been violated (e.g., since the workers stopped before actually working on the wrong motor, an actual procedural violation and increase in risk from having two containment fan coolers unavailable may not have occurred). On the other hand, expected standards of performance clearly have not been met and the significance of potentiall'y causing two containment fan coolers to be unavailable meets the threshold of significance for inclusion in an inspection report.
To clarify these standards, the inspector may choose to include in the report narrative a statement such as: "In later discussions with the inspector, the maintenance supervisor stated that prejob briefings for safety-related tasks are expected to ensure that workers understand the exact nature of the work to be performed, including means of identifying the proper components involved." Similar statements might be included regarding clear communications with the control room, job-site verification of correct components, etc. In addition, it would also be appropriate to discuss 'any potential risk increase that would have occurred if the i
wrong motor had actually been worked.
Whenever possible, the inspector should seek to tie the finding to a documented program or expectatio_n (e.g., a generic communication on wrong-component or wrong-train events, a licensee's previously established self-checking program, etc.).
3.
Addressina the Need for Licerisee Corrective Action.
Since the standards discussed here may be in areas outside NRC requirements, they may not be used I
as the basis for rea'uestino licensee corrective action either orally or in the inspection report. When rafety issues are involved, a responsible licensee will likely take corrective actions, and these actions should be documented in the inspection report as appropriate.
If the licensee fails to take proper corrective action for a safety matter and the problem recurs or additional safety issues result, the licensee may be in noncompliance with 10 CFR Part 50, Appendix B, Criterion XVI," Corrective Action."
Finally, in extreme cases where the licensee refuses to take corrective action for a matter of immediate safety significance, the NRC may exercise its authority to impose an order, even if the licensee has not violated an existing regulation or license condition. Any such situation should result in prompt involvement by NRC management (including OE and the Office of General Counsel).
4.
Avoidina Inadvertent Backfits. 10 CFR 50.109 establishes specific regulatory authority for the NRC to impose new requirements on reactor licensees involving the addition, elimination, or modification of structures, systems, or components at operating facilities. In order to impose a backfit, the Commission must make a finding that the action will result in substantial additional protection of public health and safety or the common defense and security.
As discussed in NRC Management Directive 8.4, an NRC staff recommendation that the Commission impose a backfit should only be made after extensive deliberation and evaluation of all associated circumstances. For routine discussions of safety issues in inspection reports, care must be exercised to avoid making an inadvertent recommendation that could be construed as an NRC backfit.
0610* DRAFT Issue Date: DRAFT 5/19/99
c.
Documentina Emeroency Preoaredness Exercise Weaknesses. Esch extreiss wsekness (EW) identified by ths NRC, tnd not by ths lic:nssa, while monitoring a licansso's emergency preparedness exercise should be described in detail. 10 CFR Part 50, Appendix E, Section IV.F.5 requires the licensee to take corrective actions for identified weaknesses or deficiencies. The report narrative should be sufficiently detailed for the licensee to determine what types of corrective actions are needed. Licensee identified exercise weaknetses may be inc'uded when they meet the general criteria of Section 5.02.b for including items in a report.
d.
Documentina Manaaement Issues. Inspectors should not draw conclusions regarding licensee management effectiveness. NRC requirements related to licenses management are limited, and few inspectors have professional training in evaluating administrative or managerial skills, the appropriate level of staffing for a given licensee task or program, or similar issues.
Inspectors should seek, however, to identify and document findings that will assist NRC management in assessments. When discussing specific findings the inspector may identify specific, concrete ways in which staffing changes or management involvement have contributed to that finding.
NON-SPECIFIC: "The continued motor-operated valve deficiencies showed a lack of management support in this area."
SPECIFIC: "The licensee determined that the continued motor-operated valve deficiencies had several apparent causes, including (1) the failure to schedule outage repairs for the valves, (2) the failure to adequately track the repetitive failures, and (3).
the lack of follow-up to internal audit findings in this area."
n The dete4s given in the second example are much more meaningful.
Finally, when referencing statements made or positions taken by " licensee management,"
the inspection report should be as specific as possible as to which licensee manager or management area is being referenced (e.g., "the Unit 2 operations manager," "the director of regulatory compliance," or "the engineering manager for plant modifications").
05.06 Jreatment of Open items in Reactor inspection Rennrts. Issues that merit additional inspection are identified by a unique tracking number and entered into the IRAM system by the originating inspector or office. Open items include unresolved items, violations, deviations, non-j cited violations, NRC identified exercise weaknesses, and licensee event reports (LERs).
a.
Initiatina Ooen items. The action of initiating an open item is a commitment of future resources, and should therefore only be used when some specific licensee action is i
pending, or when needed information is not available at the time of the inspection. When the inspector believes that the additional information may reveal the issue to be a matter of noncompliance, an unresolved item should be initiated. For an unresolved item, the I
report should identify the actions or additional research needed to resolve the issue.
Issues of noncompliance (except for minor violations) and NRC identified exercise weaknesses should always be assigned an IRAM number for tracking purposes. When an inspection involves multiple violations (or multiple examples of a single violation), the inspector should be careful to ensure a one-to-one correspondence between the number of IRAM entries and the number of " contrary to" statements in the accompanying Notice of Violation. The NRC Enforcement Manual provides additional guidance on tracking and following up issues of noncompliance.
Issue Date: DRAFT S/19/99 0610* DRAFT
Upon receipt, LERs should automitically be e nt: red into tha lRAM cyst:m for tracking and follow-up.
b.
Follow-Uo and Closure of Ooen items. The level of detail devoted to closing open items depends on the nature and significance of the additional information identified. For example, the closure of an unresolved item should, at a minimum, summarize the topic, summarize the inspector's follow-up actions, evaluate the adequacy of anylicense actions, determine if a violation occurred, and include enough detail to justify closing the issue in closing out a violation, if the licensee's " Response to a Notice of Violation" already has given an accurate description of the root cause, corrective actions taken, and other aspects, and the inspector identifies no other instances of the violation, the close-out description should be correspondingly brief. Normally NCVs and exercise weaknesses will be opened and closed in the initiating inspection report.
EXAMPLE: "(Closed) Violation 999/98008-03: failure to properly post a high radiation area. The inspector verified the corrective actions described in the licensee's response letter, dated March 28,19_, to be reasonable and complete. No similar problems were identified."
c.
Treatment of Licensee Event Reports. All LERs should be followed up and given formal closure in an inspection report. However, the level of detail provided in the report will vary depending on the significance of the LER and the results of the inspector's follow-up.
Because the LER is already on the docket less discussion will be required.
For LERs involving minor issues, where no new equipment, system, or human performance problems are identified, and where the inspector's follow-up does not result in new information or additional perspectives, the LER closure should be correspondingly brief:
EXAMPLE: "(Closed) LER 999/1998-003-00: auxiliary building ventilation actuation.
This LER was a minor issue and was closed."
Most LERs relate to some aspect of equipment, system, or human performance problems.
If these probierns have already been discussed and dispositioned separately in another section of this or a previous report, the LER closure may simply consist of a reference to that discussion:
EXAMPLE: "(Closed) LER 999/1999-002-00: high pressure safety injection isolation.
This event was discussed in NRC Inspection Report 50-999/99-01. No new issues were revealed by the LER."
When the LER involves more than a minor issue, and the issue has no.1 been discussed and dispositioned in another section of this or a previous report, the LER closure should provide, at a minimura, a basic description of the event and a discussion of the safety significance of the event, as determined by the SDP analysis. The discussion should include the licensee's immedirite response and subsequent corrective actions, the root cause or causes, a summary of the inspector's follow-up actions, if any, and any required enforcement actions. The discussion should be brief and concise, except in cases where the NRC's information and perspectives differ from the licensee's information and perspec-tives described in the LER. If the inspector's follow-up does not result in new information or additional perspectives, the report should not uselessly reiterate the detailed event description from the LER.
0610* DRAFT Issue Date: DRAFT 5/19/99
Nots that LERs frequrntly involvs violations of TSs or oth:r requirsments. As with oth:r I
report findings, if ths LER is discuned in a minnar that implits a violation mty havs occurred (either as part of the event itself or in the underlying root cause), the noncompliance must be clearly dispositioned in the report as a violation, an apparent violation, or an NCV, as appropriate, or a statement included clarifying that "this event did not constitute a violation of NRC requirements."
d.
Avoidina "lmolied" Insoection Follow-Up Items.
The inspection report should not commit to future NRC attention in a particular area.
0610-06 GUIDANCE -INSPECTION REPORT FORMAT Whenever possible, NRC inspection reports should conform to the standard formats described in this section and illustrated in the attached exhibits. This standardization in format significantly enhances readability and information retrieval, which in tum increases efficiency and improves the ability to integrate inspection results. Exceptions should be made for major team inspection reports, augmented inspection team (AIT) reports, and other cases where the specifically directed focus of the inspection does not easily fit into the baseline inspection process and subtopics given in the standardized report outline.
06.01 Cover Letter. Inspection reports are transmitted using a cover letter from the applicable NRC official (branch chief, division director, or regional administrator) to the designated licensee executive. Cover letter content varies somewhat depending on whether or not the inspection identified noncompliances. In general, however, every cover letter uses the same basic structure.
NOTE: Management Directive (MD) 3.57, " Correspondence Management," Part lil provides guidance for NRC letters, including inspection report cover letters. In addition, the NRC Enforcement Manual provides standard transmittal letter formats for inspections in which noncompliances are identified.
a.
Addresses. Date. and Salutation. At the top of the first page, the cover letter begins with the NRC seal and address, followed by the date on which the report cover letter is signed and the report issued.
For cover letters transmitting reports with issues assigned an escalated action (EA) number, the EA number should be placed in the upper left-hand comer above the principle j
addressee's name.
The name and title of the principle addressee are placed at least four lines below the letterhead, followed by the licensee's name and address (see Exhibit 2, the sample report). Note that the salutation is placed after the subject line.
I b.
Subiect Line. The subject line of the letter should state the type of inspection report (e.g.,
"NRC INTEGRATED INSPECTION REPORT," "NRC REQUALIFICATION TRAINING INSPECTION REPORT") followed by the report number. The words " NOTICE OF VIOLATION" (or " NOTICE OF DEVIATION" etc.) should be included if such a notice is accompanying the inspection report.
c.
Introductorv Paraaraoh. The first paragraph of the letter should give a brief introduction,.
as follows:
EXAMPLE: "On July 24 through August 31,19__, the NRC completed a safety inspection at your facility. The enclosed report presents the results of that inspection."
issue Date: DRAFT 5/19/99 0610* DRAFT
d.
Body of the Leth Sampl3 cov r isttsr information will b3 included in the lat:st enforc: mint instructions, e.
Closina. The final paragraph consists of legal boilerplate that varies based on whether or not enforcement action is involved. The signature of the appropriate NRC official is followed by the docket number (s), license number (s), and lists of enclosures and distribution.
06.02 Cover Pace. The report cover page provides a quick-glance summary of information about the inspection (see Exhibit 2). It contains the dates of inspection, the report number, the names and titles of participating inspectors, and the name and title of the approving NRC manager.
NOTE: A record of inspector and reviewer concurrence in the report is separately recorded and maintained by the issuing office. This concurrence record is not included in the generally distributed version of the report.
06.03 Summarv of Findinas/ Plant issue Matrix (PIM). The summary should be informative but concise. An ideal inspection report summary will be useful as an overview tool for licensee management and for NRC staff.
a.
Introduction. The summary should begin with a one-or two-sentence introduction that covers the type of inspection, the scope (i.e., the licensee programs or baseline areas inspected), and any special details.
EXAMPLE: "This integrated inspection report covers a 6-week period of resident inspection, announced inspections by regional engineering and radiation specialist inspectors, and an unannounced visit by a regional safeguards inspector."
b Presentation of Sianificant Findinas. The issues thatfollowshould be listed by comerstone in the order specified Section 06.03.c.3. In addition, within each cornerstone, the items should be listed in the order of importance. Findings SHALL include the results of the SDP review; le, Green, White, Yellow, or Red. There is one exception to the requirement to list findings by comerstone. The yearly identification and resolution of problems inspection procedure requires that assessments of the corrective action program be made.
These assessments shall be included under the heading of " Corrective Action Program."
The Summary of Findinos should be compiled by scanning each report section and writing a crisp, short summary sentence or sentences for each issue of note-noncompliatices (including apparent violations) and significant findings. Not all entries in the report details need to be included in the Summarv of Findinas. The threshold of significance for including an issue in the Summarv of Findinos should normally be based on whether the issue was an SDP phase two candidate for operating reactor findings, required SDP consideration for nonreactor findings, reports findings in areas without a current SDP, such as shutdown issues, or reports a noncompliance. Other findings which document a more than minor aspect of licensee performance may be included, even if the findings do not meet SDP review criteria. In addition, programmatic findings from the annual corrective action inspection should be included, even if they are neutral. Minor violations, if they are included in the report, do not have to be included in the summary. Key observations and findings should be drawn from the report and summarized. It is not expected that the summary will be an exact copy of the words in the report details, however, inspectors should ensure that the summary is consistent with the details (exact copy is acceptable). Acronyms should not be used from previous items, because items may bo in different order in the PIM 0610* DRAFT Issue Date: DRAFT 5/19/99 l
l
NOTE: URis ns:d not be discurs:d in ths Summarv of Findinas (i.e., whara more informition is need:d to reach a finding.) How:vsr, URis may b3 includ:d whan enforcement is still under review, but the findings have been identified. Violations, NCVs, and NRC identified exercise weaknesses shall be included in the Summarv of Findinas.
The usefulness of findings will be increased by concisely stating the root cause (if the root cause has been determined).
c.
ELid All enkies in the Summarv of Findinas will be transferred to the PIM. Although the 4
PIM is not a direct part of the inspection report, instructions are included here to assist inspectors in identifying the information required for the PIM during the inspection.
1.
The PIM shall include the following information; comerstone, type, date of
- occurrence, source (normally expected to be the inspection report number), who identified the finding, SDP result, and item description. The PIM shall be sorted by comerstones and entries *vithin the cornerstone shall be listed in reverse chronological order. The PIM should contain the information from the past 12 months. To discourage comparisons of PIMs based only on the number of issues, the PIM should not include a total entry listing. More detailed instructions on PIM entries follow:
2.
ITEM DESCRIPTION: The information from the Summarv of Findinas shall be transferred to the PIM as written, except that minor editorial changes may be made; for example, removal of the paragraph number at the end of the discussion or addition of an LER occurrence date that is significantly different than the report date. In addition, enforcement-related amplifying information should be added to the end of the item description. Specifically, if enforcement discretion is granted, the applicable section of the Enforcement Policy should be included, as well as the severity level. For final escalated enforcement actions, the severity level (may also be in the TYPE column), whether a civil penalty was issued and the amount, f actors, use of discretion, etc. should be included. Apparent violations should be included in the PlM. When apparent violations and URis are resolved, the PIM entry shall be modified to represent to final resolution.
After an inspection report is issued, if is determined that the PIM entry is unclear, it may be edited appropriately to clarify the issue, with the goal of improving the understanding of issue. However, only information contained in the body of the report shall be used. Care should be taken to ensure that new or undocketed information is not inadvertently introduced into the PlM. Any changes of content shall be included within brackets, [], to clearly show the editing. Use of brackets is not required for addition of the clarifying information discussed in the previous paragraph.
3.
CORNERSTONE:
Identify and document. Leave blank for yearly corrective action program assessments.
Use the following
'l abbreviations:
lE - Initiating Event MS - Mitigating System Bl - Barrier Integrity EP - Emergency Preparedness j
OS - Occupational Radiation Safety PS - Public Radiation Safety PP - Physical Protection i
Issue Date: DRAFT 5/19/99 0610* DRAFT
1 4.
TYPE:
Id:ntify and docum:nt. Usa the following abbrsviitions:
eel-Escalated Enforcement issue VIO -
Notice of Violation (include severity level)
NCV -
Noncited Violation DEV -
Notice of Deviation ED -
Enforcement Discretion LER -
Licensee Event Report URI-Unresolved item NEG -
Negative finding LIC -
Licensing issue EW -
Emergency preparedness weakness MISC -
Miscellaneous (declared emergencies, corrective action program findings, etc.)
5.
SOURCE:
Identify and document. Normally this will'be the inspection report number listec; ay year and three digit report numbor without spaces, for example, 1999001. Multiple SOURCE codes should be used where appropriate.
If used, the most significant item should be listed first. For example, if there is a related LER, the SOURCE column should include both the inspection report and LER numbers, with the inspection report listed first.
6.
DATE:
Identify and document. For PIM entries which describe an event or significant issue that has a clear date of occurrence, use this date when i
documenting the item in the PIM. For other entries such as LERs, use the date that f
c the information source was issued; with the exception of NRC Inspection Reports, in which the last date of the inspection period should be used. When the LER.
occurrence date is significantly earlier than the report date, add the occurrence date to the item description, to put the time of the issue in context.
7.
IDENTIFICATION:
Identify and document as NRC, licensee (LIC), or self-revealing (SELF); Self-revealing refers to those issues that are identified by an occurrence or action that was not an initiative of the licensee or NRC. Examples include valve misalignments identfled during a TS required surveillance test, modification errors that are not ider.t fled until an actual system demand occurs, etc.
8.
SIGNIFICANCE: Identify and document the " color" of the issue in accordance with the SDP. Leave blank for yearly corrective action program assessments.
06.04 Tabio of Contents.
For reports of significant length (i.e., in which the report details section exceeds 10 pages), the writer should consider including a table of contents as an aid to clarity.
06.05 _ Report Details: Use of the Standardized Reoort Outline. The report details should be topically arranged in accordance with the standardized report outline, included as Exhibit 1. This does not mean, of course, that each outline topic should be covered in each report. To the extent that inspection is performed in a particular area (e.g., inspection of " gaseous and liquid effluents"),
the resulting findings should be placed in the corresponding standard section of the report (e.g.,
in 2PS1 of the standardized outline; see Exhibits 1 and 2).
NOTE: Conformity to the standardized outline should not result in artificially fragmenting an event description or separating report details that would logically be presented together. For events the discussion of the entire event shall be included in the most appropriate area. Individual findings, which result in summary of findings entries, will be 0610* DRAFT Issue Date: DRAFT 5/19/99
link:d to th3 most appropritts comirstons. R:gardirss of what s:ction the writ:r finds most appropri:ts, ths bric d tails need only b3 prn:nt:d once.
The sample report included as Exhibit 2 illustrates the effectiveness of this practice. While not all sections of the standardized report outline are covered, the report details are maintained in a coherent, predictable order that corresponds to each cornerstone, and the inspectable area procedure (s) within the comerstone.
06.06 Report Details: Internal Oraanization of Soecific Sections. Differences in the nature, significance, and complexity of individual findings results in considerable variety in how those findings are organized and presented. However, as shown in the attached sample report (Exhibit 2), the overall organization of each report section should follow the same basic progression of logic:
inspectable area, optional title, scope, and observations and findings a.
Inspection Scooe. As discussed earlier, this description should be completwd k!ual, but concise to aid the reader in putting observations and findings in context. Inspections
. which have no observations and findings should not normally be recorded in the report, with one exception. When a specific inspection has been completed in an inspection area with a limited sample size, and this information may be useful in planning future inspections, a brief description of the area (s) inspected may be included. For example, design modifications have a limited sample size, so to preclude duplication it is acceptable to simply list design modifications reviewed. However, it is not normally useful to list examples of postmaintenance testing witnessed, in that postmaintenance tests are typically.not repetitive and have a potentially unlimited sample size. In the case where an inspection scope is included for future reference only, the following sentence shall be the only entry under Observations and Findinas: "There were no findings identified and I
documented during this inspection."
b.
Observations and Findinas. This portion of each report section should be used to~present, in a concise narrative format, the inspection results. The first sentence or two of this section should provide the results of the entire discussion. This should be very brief, and unlike the Summarv of Findin_gg does not need to " stand alone," because the following discussion will provioe the supporting details. At this stage in the report, the inspector may choose to simply number issues sequentially, with appropriate subheadings, or may use another method of organizing the findings (see Section 07.02 and Exhibit 2 for additional guidance on organization methods). This section should provide the information required to make and support an SDP determination. The report should include, as necessary, a statement of the observation (s), resultant finding (s), assumptions made, duration, mitigation, accident scenarios, and worst case safety significance to a comerstone(s) from any increased risk. When discussing accident scenarios and worst case safety significance, clearly indicate if the condition occurred or only could have occurred. When i
a report discusses more than one issue under a particular inspectable area procedure, the most important issue shall be discussed first.
During performance-based inspections, inspectors are going to observe compliance-related issues which will immediately be recognized to be within the licensee's response band (green), without a formal SDP determination. For these types of issues, inspectors should no.1 collect and report the background information required for a SDP candidate.
Instead, inspectors should simply state that the issue did not meet any of the criteria for an SDP review. The issue should then be reported with only enough detail to support the finding (s).
06.07 Exit Meetina Summarv. - The final section of each reactor inspection report should be a brief summary of the exit meeting. This summary normally should include the following elements:
lasue Date: DRAFT 5/19/99 0610* DRAFT
c.
Characterization of Licensee Response. In gin:ral, tha r: port should not chiracterize a licensee's exit mnting r:spon 3 cs ons of whol:hiirt:d ccc:ptance of ths insp:ction findings, if the licensee generally agreed with the findings presented, the exit meeting characterization might read as follows:
EXAMPLE: "The inspectors presented the inspection results to members of licensee management at an exit meeting on June 12,19.__. The licensee acknowledged the findings presented."
On the other hand, when the licensee disaarees with the inspectors' finding, this position should be briefly and specifically characterized (e.g., "the plant manager stated that he believed the violation of TS 4.3.1.2, regarding a reactor trip system surveillance, to be of no safety significance"). Specific items discussed elsewhere in the report should not be described in this section in detail.
b.
Licensee Oral Statements and Reaulatory Commitments. If, at the exit meeting or at any other time during the inspection, the licensee makes an oral statement that it will take a specific action, the report should attempt to accurately characterize that statement. As determined by the significance, complexity, subject area, and resource expenditure involved, the inspector should ensure that such oral statements are made or endorsed by the proper member of licensee management. Inspectors should be careful to differentiate between (1) licensee general descriptions of "vo'untary enhancements" or general intent; and (2) oral statements of the licensee's intent to make a specific regulatory commitment
- (i.e., to submit, on the docket, a written commitment to take a specific action).
Because regulatory commitments are a sensitive area, the inspector should also ensure that any reporting of such a licensee oral statement is accurately characterized. To-ensure a clear mutual understanding of such issues, when the licensee makes an oral statement reflecting the intent to make a regulatory commitment, the report issuing office may wish to restate, in the report cover letter, the NRC's understanding of that proposed commitment, and ask the licensee to clarify any differences in understanding.
c.
Absence of Proorietary Information. At the exit meeting, the inspectors should verify whether or not the licensee considers any materials provided to or reviewed by the inspectors to be proprietary.
NOTE: When an inspection is likely to involve proprietary information (i.e., based on the technical area or other considerations of inspection scope), the topic of how to handle such information should be discussed at the entrance meeting.
)
If the licensee does not identify any material as proprietary, the exit meeting summary should include a sentence te
.,e effect (see inspection Manual Chapter (IMC) 0611 on actions to take if the report '1ch des proprietary material).
EXAMPLE: "The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified."
d.
Subsecuent Contacts or Chanaes in NRC Position. The report writer should briefly discuss any significant contacts between the inspectors and licensee staff or management that occur after the exit meeting (e.g., to discuss new information relevant to an inspection finding), in addition, as discussed earlier, if the NRC's position on an inspection finding changed significantly after the exit meeting, that change should be discussed with the licensee before the report is issued.
0610* DRAIT Issue Date: DRAFT 5/19/99
06.08 Reoort Attachments. Tha attachments discussed b: low chould b3 included at ths end of ths insp!ction rtport.
a.
Partial List of Persons Contacted. The report writer should list, by name and title, those individuals who furnished significant information or were key points of contact during the inspection (except in cases where there is a need to protect the identity of an individual).
An exhaustive list is neither required nor desirable; 5 - 10 key individuals normally is sufficient. The alphabetized list should include the most senior licensee manager present at the exit meeting. The list should also include other NRC technical personnel who had significant involvement, if they were not listed as inspectors on the cover page.
b.
List of items Opened. Closed. and Discussed. The report should provide a quick-reference list of items opened and closed, including the item number, the IRAM code for the item, and a brief phrase (10 words or less) t.% scribing the item. Open items that were discussed (but not closed) should also be included in this list, along with the report section number in which the items were initiated. See the sample list included with Exhibit 2.
c.
List of Baseline insoections Performed. The report should include a listing of all the inspectable area inspections performed. The fol!owing two sentences shall also be included at the top of the list, "The following inspectable area procedures were used to perform inspections during the report period. Documented findings are contained in the body of the report."
4 d.
List of Acronyms. Reports of significant length (i.e., in which the report details section exceeds 10 pages) should generally include a list of acronyms as an attachment. For reports in which a relatively small number of acronyms have been used, such a list should be considered optional. In all cases, however, acronyms should be clearly defined when
(
first used in text, regardless of whether a list of acronyms follows the report narrative.
)
l 0610-07 GUIDANCE -INSPECTION REPORT STYLE A package is frequently judged by its wrappings; similarly, an inspection is often evaluated based on the clarity of the inspection report. The best engineer, putting forth the most intensive inspection effort, may not succeed in delivering the desired safety messages if the inspection results are shoddily presented. On the other hand, a clear writing style that communicates the desired safety issues increases the professionalism of the product and can enhance the credibility of a good inspector (and of the agency).
07.01 Peculiarities of Govemment Technical Writina. Govemment technical writing, as it appears in NRC inspection reports, combines aspects of scientific, legal, and corporate communication.
Since few inspectors possess the combined skills of a scientist, attomey, and business executive, certain aspects of NRC report writing have been standardized., Most new inspectors find these standardized aspects peculiar, and some writers adapt more easily than others. Four specific writing style features will be discussed here: using "boilerplate"; writing in standard grammatical tenses; avoiding " purple" language; and using technical, legal, or local jargon.
a.
Boilerolate. The first aspect is the use of standard paragraphs or phrases, called " boiler-plate," for certain sections of the inspection report. Boilerplate is provided in various NRC references as a writer's guideline, (1) to achieve consistency throughout the agency among writers whose individual styles may vary, and (2) to provide precise wording for statements that may have legal implications (such as the standard legal or enforcement wording that appears in cover letters and NOVs). Used properly, boilerplate can save time and effort.
Issue Date: DRAFT 5/19/99 0610* DRAFT
Boilirplits thould np.1 be used Es tha cubstituts for report contant, such cs whan discussing inspection m:thods, activiti:s insp:ct:d, or insp:ction findings (ses S:ction 05.03.c). On the other hand, once effort has been expended to develop clear standards in a given inspection area, the inspector may find it useful to save and repeat those
' statements in later inspections of the same or similar areas.
b.
fa.rammatical Person. Tense. and Voice Used in NRC Insoection Reoorts. The use of specific grammatical conventions is another style peculiarity of NRC report writing that some inspectors initially find awkward. The narrative sections of a report should be written in the third person, in the past or past perfect tense, in predominately active voice, as follows:
l 1.
Always write in the third person.
RIGHT: "The inspector watched the mechanic remove the check valve."
WRONG: "I watched the mechanic remove the check valve."
2.
In general, use the past or past perfect tense.
RIGHT: "The plant manager stated that the review committee had been fully staffed as of April 19_."
WRONG: "The plant manager states that the review committee was fully staffed as of April 19_."
NOTE: As a rule of thumb, use the past tense (e.g., "the pump cavitated..." OR "the pump was cavitating...") when writing about events that occurred during the inspection. Use the past perfect tense (e.g., "the pump had cavitated..." OR "the pump had been cavitating...") when writing about events that occurred before the inspection.
EXCEPTION: When quoting or paraphrasing existing literature, the present tense may be used (e.g., "10 CFR 50.71 states..."). However, if using the present tense when quoting a licensee procedure, be certain that the quotation is current at the time of issuing the inspection report (or give the date and revision number of the procedurrs from which the quote was taken).
3.
Use predominately active voice (subject-verb-object). Passive voice tends to be more wordy and indirect, less interesting, and thus more difficult to read and comprehend.
ACTIVE: "The auxiliary operator reported that lube oil pressure was rising."
i PASSIVE: "It was reported by the auxiliary operator that lube oil pressure was rising.
in addition, passive voice sometimes fails to identify the subject, or the performer of the action.
PASSIVE: "It was reported that lube oil pressure was rising" OR " Lube oil pressure was reported to be rising."
in cases such as the latter example, the use of passive voice actually results in omitting information that could be important in evaluating the significance of the observation.
0610* DRAFT Issue Date: DRAFT 5/19/99
c.
Avoidina Purole Lanauaae. 'Purpl3" langurga_ refsrs to words or phrases thit h:va und:sirrbla connotitions or hidd:n implications, such that a read:r might misconstrus the writer's meaning. Several examples follow:
EXAMPLE: " Foreign material exclusion controls had been carelessly disregarded by the maintenance staff."
PROBLEM: Because terms such as " careless disregard," " willful," and " deliberate" represent agency conclusions with specific NRC enforcement connotations, they should not be used in a report narrative unless they represent a formal agency conclusion.
EXAMPLE: "By waiting until the last minute to return the charging pump to service, the licensee narrowly missed getting an NRC violation."
PROBLEM: The inspection report is not a forum for the inspector's opinions or conjecture; observations should be factual rather than speculative.
d.
Usina Technical. Leoal. and Local Jaraon. Because of the specialized technical and legal aspects of NRC regulation, inspectors must maintain sensitivity to the use of specialized vocabulary in inspection reports. The use of technical and legal jargon is expected and necessary; however, inspectors should explain terms that are likely to be unfamiliar to the
" target reader" (see discussion in Section 05.03.a).
In addition, inspectors should avoid the use of " local jargon," terms or phrases that have specific meaning for a particular plant or region of the country but are not widely understood. Examples include terms and phrases such as " tailgate" (when used to refer to an informal meeting), " work-around," and " command and control." Because these labels are used routinely by a localized group, inspectors may become accustomed to their use and assume that they have a consistent, widely understood meaning. When used in inspection reports, however, readers from another facility or region may be given an unclear or unintended meaning. To avoid miscommunication, report writers should define such terms and phrases whenever used or, if possible, use an altemate word that has a clear, dictionary-defined meaning.
For example, consider the following usage of the term " work around":
1 ACCEPTABLE:
"The licensee's failure to resolve longstanding concems has I
contributed to an excessive reliance on operator " work-arounds." The term " work-around," as used in this report, refers to nonroutine actions performed by the operating crew to compensate for equipment not functioning as designed."
In this case, the writer chose to use local jargon by including the term " work-around." By defining the term immediately after its initial use in the report, however, the writer avoided the possibility of vagueness or an incorrect interpretation for a reader not familiar with the term.
NOTE: The example above would npf be acceptable if the writer had omitted the second sentence.
As an alternative, the writer may choose not to use the term at all:
ACCEPTABLE: "The licensee's failure to resolve longstanding concems has contributed to an excessive reliance on the performance of nonroutine actions by the operating crews, to bypass or compensate for equipment not functioning as designed."
07.02 Clear Oraanization. The attribute most likely to make a report appear unprofessional is a lack of organization. This writing style problem is evidenced in reports that (1) leave the reader issue Date: DRAFT 5/19/99 0610* DRAFT
uncertain as to tho m in points or ov r:Il m:ssagn; (2) blur ths distinctions b:twsen obs rv:tions, and findings; (3) fail to pr:s:nt findings in t:rms of r:lativs significanes (i.e., trsat all findings with I
the same priority and level of detail); (4) present redundant information throughout the report (e.g.,
use the same observation to create findings in multiple areas); and/or (5) make contentions in the cover letter or summary of findings that do not match the report findings.
Using a standardized report outline and a standardized format for report sections helps writers to
' achieve more coherent report organization, but is frequently not enough. For writers that struggle I
with organization, or for good writers that want to further improve their writing style, several specific actions can be of use.
l a.
Manaaina the Writina Process. _Many writing courses are designed to turn out an effective business memo or convincing letter. These courses teach the mid-level executive how to break the writing process down into manageable steps. The same techniques, slightly modified, can be applied to an NRC inspection report.
i Writing involves planning, brainstorming, organizing, drafting, and revising. Trying to perform all of these activities at once will produce " writer's block," incoherent presentation of findings, sloppy final drafts, and extended cycles of report review and revision. A more effective writer breaks the writing process into stages to avoid frustration, save time, and produce a more professional product.
Appendix B provides detailed guidance on managing the writing process in this manner.
b.
. Standard Oraanizational Techniaues. Within the " Observations and Findings" portion of each report section, the inspector must determine how to organize the findings relevant to that area. When few findings exist, organization is simple. However, when an extended evaluation has resulted in a large number of findings, the inspector must group the
)
findings into logical categories, and arrange the findings in each category into some logical j
order. The following methods may be used:
'e Order of Imoortance. Within each inspectable area, present the observations and findings beginning with the most significant.
- Findina Results. Within each Observation and Findinas the results of the entire discussion should be provided in the first sentence (or two sentences) to hi-lite to the reader the most important aspect of the discussion that will follow.
- Chronoloaical.
This method is useful in presenting the essential details of a complicated event.
- Comoarison/Contrasti This is the organizational method used for writing NOVs, in which the regulation is quoted or paraphrased and closely parallel language is used in the " Contrary to" statement. This is also an effective method of presenting a negative finding in the report details, c.
Use of Repetitive Formats. Most inspection writing tasks are repetitive in organization.
By identifying the basic structure of each type of report section, the inspector can use this repetitiveness to make the organization of details much easier.
EXAMPLE: Consider the details usually present in closing an open item. List the types of information that may be included: a description of the issue; licensee corrective actions; an indication of the item's open or closed status; the applicable regulatory criteria; the inspector's actions to verify resolution of the issue; the root cause (when i
following up on an event or a violation); and the IRAM number of the open item.
j i
0610* DRAFT Issue Date: DRAFT 5/19/99
Now errzngs th:se pieces of inform: tion into a logical ordar, and crsita a ssriss of ct ps for documsnting ths closura of an open it m:
- 1. Give the IRAM number of the open item.
- 2. Indicate next to the number whether the item will rentain open or closed.
- 3. Briefly describe the issue.
- 4. In the description, reference the regulatory criteria, if applicable.
- 5. State the root cause,if desired.
- 6. Describe the licensee's corrective actions.
- 7. Describe the actions the inspector took to verify resolution of the issue.
Now examine the examples of follow-up items given in the sample report (Exhibit 2).
j in each case, the narrative follows the same logical pattern described above. Some types of information, such as a statement of root cause, will not be included for less significant issues or when the follow-up results are routine and straightforward. The underlying organization, however, will remain the same.
This technique can be applied to most sections or subsections of the inspection report.
Once the structure has been established and repeated several times, the organization of i
similar details in later reports can become almost automatic, saving time and effort.
07.03 Effective Revision. Because of deadlines and other pressures, many writers hand in their products for review and approval immediately after drafting, without taking the time for effective revision. Whether reviewing one's own report or that of someone else, the following guidance will i
be helpful in (1) ensuring that the report communicates clearly, (2) achieving professionalism in the writing style, and (3) avoiding embarrassi^ nistakes.
1 NOTE: The guidance below is summarized in a ' report review checklist, included as' Appendix C to this chapter.
Levels of revision should be prioritized. Some corrections are more important than others; if the reviewer begins by focusing on spelling and punctuation, he or she may be too distracted to notice larger organizational problems. The most effective practice is to revise in several stages.
The first stage is a rapid read-through to check for overall coherent organization and level of detail in each report section. If no major revisions are needed, the second stage review is then per-formed, reading to assess paragraph structure, sentence style, and clarity of syntax. The final stage is a careful word-by-word proofread to check for spelling, punctuation, and accuracy.
a.
Hiahliahtina the Messaae. The first stage review ensures that the most significant findings will be evident to the reader.
- Are the main topics of the cover letter supported by findings in the Summary of Findinas?
- Are the findings in the Summary of Findinas consistent with the report details?
- Are the main ideas in each section clearly developed (observations translated to findings)?
- ls the level of detail appropriate for the significance of the finding and in keeping with the technical expertise of the target reader?
b.
Checkina the Style. Even a well-organized report can miscommunicate. Govemment technical writers sometimes lapse into a "bureaucratese' style that can obscure meaning.
Issue Date: DRAFT 5/19/99 0610* DRAFT
r"'
Long piragr:phs and s:ntanc:s, pompous linguigs, and usslass phrasts cre symptoms of an und:sirable burseucratic style.
To illustrate: -imagine sitting in a classroom, listening to the lecture of an eloquent instructor who knows the material well, who has an obvious enthusiasm for the task, who genuinely wants the students to leam, who never loses the thread of thought, who always i
has the appropriate answers, but whose fatal flaw is neglecting to give breaks, and whose effectiveness is therefore severely reduced by never giving the students a chance to pause, to breathe, to digest the material. Long sentences (and long paragraphs) have a similar effect. (To understand this effect, use a single breath to read aloud the first sentence of this paragraph.)
By contrast, the second-stage revision can enhance report clarity, in part, by noticing the amount of " white space" on the page. Where appropriate, add emphasis techniques, such as subheadings or lists. Make each paragraph coherent and concise. Ensure the predominate use of active voice. Eliminate wordy phrases, false subjects, and other symptoms of "bureaucratese" that weaken sentence clarity. A crisp, clear style will add far more credibility to a report than an inflated vocabulary.
c.
Proofreadina. Unfortunately, there is no good substitute for a final, word-by-word proofread. Computer spell-check systems are helpful, but will not identify cases in which the wrong word is used for the context (e.g., " corrective actions were directed by the cite maintenance manager"). This final revision should look for such errors as:
- Improper subject-verb agreement (and other grammar errors),
- Undefined acronyms, l
- -Typographical errors that result in the wrong word for the context,
- Missed or incorrect punctuation, and
- Metrication mistakes: errors in converting Si to/from English units.
In addition, before the final draft is released, verify the accuracy of all numbers '(including page numbers), dates, distribution, and titles.
l 07.04 Writina Stvle Guides and Useful NRC Referg.n_qqs. Few writers can recite all the rules of j
l English grammar, and few inspectors can remember the large assortment of NRC boilerplate and I
guidance. Most inspectors will find it useful to be familiar with the following references:
1 a.
The NRC Enforcement Manual provides specific guidance and boilerplate for writing associated with enforcement actions.
b.
The U.S. Govemment Printina Office (GPO) Style Manual is a reference on government writing style, covering a range of topics from capitalization to compound words.
c.
NUREG-1379, the NRC Stvle Guide. establishes specific guidance for the agency on the use of abbreviations, capitalization, punctuation, in-text references, and so forth. The i
NRC Stvie Guide is consistent with the GPO Style Manual on most matters.
d.
In addition, familiarity with a desktop writing guide adds efficiency and confidence in achieving a clear communication style. Three excellent examples follow: (1) Handbook of Technical Writina (by Brusaw, Alred, and Oliu, St. Martin's Press); (2) The Elements of 0610* DRAFT Issue Date: DRAFT 5/19/99
Ebdf (by Strunk cnd Whits, Macmillin); and (3) The Shiolev Style Guide (publi:hrd by
' Shipisy Assoclitas).
0610-08 RELEASE AND DISCLOSURE OF INSPECTION REPORTS, AND ASSOCIATED DOCUMENTS 08.01 General Public Disclosure and Exemotions. Except for report enclosures containing exempt information, all final inspection reports will be routinely disclosed to the public. IMC 0611,
" Review and Distribution of Inspection Reports," describes the various types of exempt information.
IMC 0620, " Inspection Documents and Records," provides guidance on acquisition and control of NRC records, including inspection-related documents.
08.02 Release of Investiaation-Reiated Information a.
When an inspector accompanies an investigator on an investigation, the inspector shall not release either the investigation report nor his or her individual input on the investigation report. This information is exempt from disclosure as provided by 10 CFR 9.5, subject to determination by OI. Ol reports of investigations, while in preparation or review, will not be circulated outside NRC without specific approval of the Chairman (Ol Policy Statement 23).
b.
Generally, NRC technical and safety concems can be communicated to a licensee without revealing that an investigation is contemplated or underway. However, when information cannot be released without risk of compromising an investigation, the regional administrator (RA) will inform the Director, 01, in advance that safety concerns require releasing to the licensee information related to an open investigation. The Director,01, will review the information to be released and advise the RA of the anticipated effect on the course of the investigation.
The RA will release the information only after determining that the safety concems are significant enough to justify the risk of compromising the pending investigation and any potential sequent regulatory action. Conversely, when the RA decides, after consultation with 01, to delay informing the licensee of an issue, the RA should document this decision, including the basis of determining that the delay is consistent with public health and safety considerations. Any such decision should be reexamined every three months to assure its continuing validity (see March 2,1987 memorandum from the Executive Director for Operations (EDO) to office directors and regional administrators).
c.
When an emergency or significant safety or secarity issue appears to require immeuiate action, NRC employees, at their discretion, may discuss with, show to, or provide the licensee any pertinent material they believe the circumstances warrant. If time permits, regional management should be consulted first.
An emergency situation meeting this criteria is one in which, in the opinion of the senior
'NRC employee cognizant of the situation, a present danger to public health or safety or to the common _defens,e and security requires the release of investigative information to a licensee without the delay necessary to consult with appropriate 01 personnel (see March 2,1987 memorandum from EDO to office directors and regional administrators).
d.
If an issue disclosed during an inspection is to be referred to 01 for possible investigative action, the inspection report should not contain information that would lead a reader to conclude or infer that an investigation may be opened. In this case, the report should contain only relevant factual information collected during the inspection. The referral to issue Date: DRAFT 5/19/99 0610* DRAFT
01 should be made by separ-t3 c:rrespondence, with any tdditional information nxd:d to cupport tha r:f:r-1.
END i
Exhibits 1 - 2 1
Appendices A - D i
'l 4
i i
e i
i a
i i
i 0610* DRAFT Issue Date: DRAFT 5/19/99
STANDARD REACTOR INSPECTION REPORT OUTLINE Exhibit 1 Cover Letter i
Cover Page Summaryof Findings Table of Contents (optional)
Report Details:
1 REACTOR SAFETY q
Initiating Events / Mitigating Systems / Barrier Integrity [ REACTOR - R]
Adverse Weather R01 Changes to License Conditions -
R02 Emergent Work R03 Equipment Alignments R04 Fire Protection R05 Flood Protection R06 Heat Sink Performance R07 Inservice inspection R08 Inservice Testing R09 Large Containment Valves R10 Licensed Operator Requalification R11 Maintenance Rule implementation -
R12 Maintenance Work Prioritization R13 Nonroutine Plant Evolutions R14 Operability Evaluations R15 Operator Work-Arounds R16 Permanent Plant Modifications R17
[ Deleted]
R18 Postmaintenance Testing R19 Refueling and Outage R20 Safety System Design R21 Surveillance Testing R22 Temporary Plant Modifications R23 Emergency Preparedness [EP]
Drill, Exercise, and Actual Events EP1 1
Alert and Notification System EP2 i
Emergency Response Organization Augmentation EP3 Emergency Action Level Revisions EP4
- 2. RADIATION SAFETY Public Radiation Safety [PS)
Gaseous and Liquid Effluent' PS1 Radioactive Material Shipping PS2 j
Radiological Environmental Monitoring PS3 1
. Occupational Radiation Safety [OS)
Issue Date:
DRAFT El-1 0610*:
Exhibit 1
Access Control OS1 ALARA Planning cnd Controls OS2 Radiation Monitoring Instrumentation OS3 Radiation Worker Performance OS4
- 3. SAFEGUARDS Physical Protection [PP]
. Access Control PP1 Access Authorization PP2 Response to Contingency Events.
PP3
- 4. OTHER ACTIVITIES [OA]
Identification and Resolution of Problems OA1 PI Verification OA2 Event Follow-up OA3 Other 0A4 Meetings, including Exit OAS NOTE: Any findings related to the Performance Indicator (PI) Baseline inspection shall be included under Other,40A2. In addition, findings associated with identification and Resolution of Problems or Event Followup which cross cornerstones shall be included in Section 40A of the report. [PIM i
findings wiil still require assessment as to which Comerstone(s) the finding applies.] In addition to the baseline inspections, supplemental inspections may be required during the pilot. Results of these nonbaseline inspections may be included under the "other" category. However, to protect j
the identification of allegation-initiated issues, inspectors should try and include allegation follow-up I
discussions in the appropriate areas of sections 1 through 3 of the report. LERs which are determined to be minor can be closed under Other,40A4.
)
0610*:
Exhibit 1 El-2 Issue Date:
DRAFT
SAMPLE REACTOR INSPECTION REPORT Exhibit 2 NOTE: the inspection report that follows is based on a fictional reactor licensee and a fictional inspection. The report contains realistic issues; however, any resemblance to an existing facility or actual events is coincidental.
This exhibit may be used as a sample or model report for matters of format and style. It illustrates how to use the standardized inspection report outline, and adheres to the expected internal organization for each re.nort section (as discussed in IMC 0610).
The s' ample report assumes that the SDP has been issued and is available for use in at least some of the comerstones. Therefore, the sample report does not include detailed descriptions of the how the SDP works but simply refers directly to already issued SDP tables and guidance. To the extent that any SDP is used that has not been formally issued on the docket, then that particular SDP and detailed supporting discussions will have to be included in the report.
The sample report discusses two issues which require an operating reactor SDP Phase 2 screening. One of these examples is divided into subheadings and contains a Summary section within the body of the report. The other SDP example does not contain many subheadings.
Inspectors may choose either method. Use of subheadings is recommended for longer discussions. If used, the Summarv should closely parallel the information in the Summary of Findinas in several ways, however, it departs from expected practice:
1.
This report illustrates various methods of report organization that would be appropriate to various types of inspection (e.g., events, corrective action reviews, observations of work).
Technicalissues discussed are drawn from both BWR and PWR technology.
2.
In terms of report content, the sample report illustrates the use of ' observations and findings," " thresholds of significance," varying levels of detail, and other concepts described in IMC 0610. However, the content included in this report should not be used as a ' standard" in the sense of how individual findings are treated (i.e., the fact that a particular event is described in a particular way in the sample report does not dictate that all similar events be given a similar level of detail in other reports).. As discussed elsewhere, judgments about inspection report content must be made based on the circumstances of an individual inspection, and will therefore vary.
3.
Pages are numbered continuously through this exhibit. Inspection reports should use separate page numbering for the cover letter, summary of findings, and report details.
= 4.
The report contains an issue which is determined to be within the increased regulator response band. This discussion is longer than what would be expected for report sections which discuss issues within the licensee response band.
Issue Date:
DRAFT E2-1 0610*:
Exhibit 2 SAMPLE REACTOR INSPECTION REPORT August 14,1999 Ms. Joan A. Doe, Vice President, Nuclear Greckenshire Power & Light 721Y Brick Road Stone Towers, WF 44632 i
SUBJECT:
NRC INTEGRATED INSPECTION REPORT 50-998/99-07,50-999/99-07 i
Dear Ms. Doe:
i On July 24,1999, the NRC completed an inspection at your Dirojac 1 & 2 reactor facilities. The enclosed report presents the results of that inspection.
]
Based on the results of this inspection, the NRC has determined that three [ Specific guidance for wording of baseline inspection program NOVs is being developed and will be incorporated into this sample cover letter when it is issued) These NCVs are described in the subject inspection report. If you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control. Desk, Washington DC 20555-0001; with a copies to the Regional Administrator, Region
- ths Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Dirojac facility.
[ Specific guidance for wording of baseline inspection program apparent violations is being developed and will be incorporated into this sample cover letter when it is issued]
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.
Sincerely, Samuel A. Johnson, Director Division of Reactor Projects Docket Nos.: 50-998,50-999 License Nos: XXX-77, XXX-79 Enclosure (s):
Inspection Report 50-998/99-07,50-999/99-07 cc w/ encl: L. Collinsworth, Compliance Manager R. Littlestaf, General Manager, Technical Services J. Bradwood, Plant General Manager
]
F. Buckfuller, General Counsel
. D. Soapstone, Operations Manager 0610*: Exhibit 2 E2-2 issue Date: DRAFT l
l
2 Issue Date: DRAFT E2-3 0610*:
Exhibit 2
U.S. NUCLEAR REGULATORY COMMISSION REGION X Docket Nos:
50-998,50-999 License Nos:
XXX-77, XXX-79 Report No:
50-998/99-07,50-999/9S-07 Licensee:
Greckenshire Power & Light (GP&L) l Facility:
Dirojac Generating Station, Units 1 & 2 1
Location:
11555 Granite Blvd.
Stone Towers, WF 44632 1
I Dates:
June 11 - July 24,1999 J
Inspectors:
A. Rand, Senior Resident inspector M. Heidegger, Resident inspector J. Locke, Senior Radiation Specialist P. Sappho, Reactor Projects inspector Approved by:
E. Tudor, Chief, Projects Branch 2 Division of Reactor Projects
\\
0610*: Exhibit 2 E2-4 issue Date: DRAFT
SUMMARY
OF FINDINGS Dirojac Generating Station, Units 1 & 2 NRC Inspection Report 50-998/99-07,50-999/99-07 The report covers a 6-week period of resident inspection, announced inspections by a regional radiation specialist and a regional projects inspector, initiating Events Green. The inspectors identified a noncited violation for failure to insure nondestructive examination contract inspectors were qualified. The inspector performing the core shroud inspections was not qualified. A different inspector reperformed the core shroud inspection and did not identify any weld cracks (Section 1R07).
Green. During plant startup operators failed to initiate emergency feedwater, resulting in an uncomplicated unit trip. All mitigation system remained operable and barrier integrity was not challenged. The inspectors identified a noncited violation for inadequate procedures (Section 1R14.2).
Mitigating Systems White. The inspectors identified an apparent violation of Technical Specification Limiting Conditions for Operation 3.5.2 in Unit 2 as a result of both trains of the emergency core cooling system being unavailable for approximately 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> and Train A being unavailable for approximately 18 days during power operations. The simultaneous unavailability of both trains resulted in the total loss of a mitigation i
function necessary to prevent core damage in the event of a loss of coolant accident (LOCA). The Train B unavailability was planned. However, the Train A unavailability, caused by mechanical failure of a line starter providing power for a Train A containment emergency sump outlet valve, occurred before the Train B unavailability and was not discovered until later. During that time, Train A would not have functioned following a recirculation actuation signal. NRC Staff calculations indicated an increase of 1.7E-5 in core damage probability. Based on Tables 1 and 2 of the Significance Determination Process the NRC staff determined that the screening for a small break LOCA was within the increased regulatory response band (white); low frequency and medium likelihood (E) with recovery of one train. The screening for a medium break LOCA was also within the
- Increased regulatory response band; lower frequency and medium likelihood (F) with no mitigation (Section 1R14.1).
Green. Three of four sources of cooling for the coolant charging pumps (CCPs) were unavailable for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> because of poor work planning. Cooling to the CCPs was provided by two trains of component cooling water (CCW), with a backup i
supply from two nonsafety-related service water pumps (SWPs). The SWPs were unavailable for two weeks. During this time, one train of CCW was taken out of service for heat exchanger cleaning. The staff determined that the highest contribution to core damage probability (CDP) from loss of the operating CCW train was loss of cooling to the CCPs, followed by loss of reactor cooling pump seal cooling and subsequent seal failure. The NRC staff calculated that the increase in CDP was small. In addition, the licensee later determined that operators could have restored the second train of CCW within the time calculated for seal failure.
Issue Date: ' DRAFT E2-5 0610*:
Exhibit 2
Th3 insp:ctors consider:d that ths lic:nsic's cytluition was ccc:ptibl3. Bared on Ttble21 cnd 2 of the significanca det:rmination proc:ss ths NRC stiff determined that even without credit for CCW restoration this issue was within the licensee's response band (green); medium likelihood and low exposure time (E) with two trains of intermediate head safety injection pumps for mitigation (Section 1R04).
Green. The inspectors identified that the licensee's in-progress corrective actions for failure of a drywell fan did not include resolution of the subsequent increase in drywell temperatures above Final Safety Analysis Report limits for drywell snubbers. The licensee subsequently determined that the snubbers were always functional, but that their qualification life was reduced by one year (1 R03).
Green. The inspectors identified a noncited violation in which a Unit 1 control rod was returned to service following maintenance without a required retest. The subsequent retest was satisfactory (Section 1R19.1).
Occupational Radiation Safety Green. Radiation protection technicians failed to remove all the tools and other material with low levels of radioactive contamination prior to release of a trailer as a temporary radiological protected area.. The I;censee had recently identified two i
similar release problems on radiological problem reports (Section 2OS4).
r I
0610*: Exhibit 2 E2-6 issue Date: DRAFT
Report Details l
S.
REACTOR SAFETY 1R03 Emeroent Work a.
Insoection Scope The inspectors reviewed the licensee's actions to resolve failure of Unit 1 Drywell Fan D1.
' b.
Observations and Findinas The inspectors identified that the licensee's in-progress corrective actions for failure of a drywell fan did not include resolution of the subsequent increase in drywell temperatures above final safety analysis report (FSAR) limits for drywell snubbers. The licensee subsequently determined that the snubbers were always functional, but that their qualification life was reduced by one year, On June 4,1999, with Unit 1 at full power, Drywell Fan D1 tripped. Drywell temperatures rose to approximately 230 *F and stabilized. The inspectors determined that the other drywell fan was operating and observed that the licensee immediately suspended any work which could have affected the operating fan. The licensee determined that a secondary contact in the circuit breaker for Fan D1 had failed. On June 8, the licensee replaced the contact, completed a postmaintenance test (PMT) and restored Fan D-1 to service. Drywell temperatures stabilized at 180 *F.
The inspectors reviewed the licensee's work packages associated with Fan D1 failure, which were still open for required shift supervisor and quality assurance reviews, and observed that the licensee had not identified that the drywell temperature had exceeded the FSAR limit of 200 *F for drywell snubbers.
The inspectors discussed the drywell temperatures with a shift supervisor, who issued a corrective action request to document and resolve exceeding the FSAR design temperature. In later discussions, the licensee's Quality Assurance manager stated that he considered his staff would have identified the FSAR temperature problem, because the checklist for quality assurance review of completed work packages included a check for compliance with the FSAR. Because the work package was still open, the inspectors considered that no violation of NRC requirements had occurred.
The licensee determined that the increased temperature did not affect the functionality of the snubbers, but did reduce their qualification life from six years to five years. After discussions with an NRC senior reactor analyst (SRA), the inspectors considered that the loss of the fan and increased drywell temperature did not significantly increase the risk of containment failure in response to any initiating events. Therefore, since the snubbers remained functional, this issue was determined to be within the licensee's response band (green).
1R04 Eauioment Alianmen1 One Coolina Water Source for the CCPs Issue Date:
DRAFT E2-7 0610*:
Exhibit 2
a.
Inspection Scooe During plant status review, the inspector leamed that three SWPs were unavailable.
After review of the impact of this information, the inspectors checked equipment alignments within the component cooling water system.
b.
Observations and Findinas Brief Overview Three of four sources for cooling for the CCPs were unavailable for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> because of poor work planning. Subsequent NRC and licensee review determined that the issue was within the licensee's response band bast,d on both adequate mitigation and cooling system restoration.
Discussion Individual status of the SWPs was as follows:
A - in service, supplying service water needs B - valved out of service because of a leaking check valve C - out of service because of high vibration D - out of service because of impeller failure and oil seal replacement Pumps B, C, and D had each been out of service for approximately 2 weeks.
The service water system is common and is not safety-related. Pumps A, B, and C are motor driven; pump D is diesel driven. The system provides water for auxiliary cooling, drinking, sanitary use, and building services.
In addition, SWPs B and D had a safety-related function which the licensee designated as safety significant in the maintenance rule program; these SWPs could be aligned to provide cooling for the CCPs. Either of two trains of CCW provided normal cooling to the CCPs.
The inspectors reviewed the licensee's control room logs, walked down the CCW portion of the main control board, and inspected the alignment of a portion of components in CCW train A. Based on this inspection, the inspectors determined that both trains of CCW were currently available.
The inspectors reviewed the control room logs for the past two weeks and found that CCW train A was available during the entire time. However, CCW train B was unavailable for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> because the associated CCW heat exchanger primary side was opened for cleaning of macro-fouling. During this 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> period, only one source of cooling was available for cooling the CCPs.
The inspectors and an NRC SRA reviewed the licensees Individual Plant Examination (IPE) and determined that the highest contribution to CDP from loss of the operating CCW train was subsequent loss of the CCPs and potential reactor coolant pump seal failure. The IPE indicated that reactor coolant pump seat failure would lead to a LOCA in about 90 minutes. The IPE showed that this scenario represents 12 percent of the total CDP or 4.6E-5. The SRA calculated that with only one train of cooling available to the CCPs for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, the average CDF was 8.8E-4/ year. With all four CCP cooling sources available during the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, the average CDF would have been 1.6E-6/ year.
0610*: Exhibit 2 E2-8 issue Date: DRAFT
This repr:s:nted an incr:cse of 8.8E-4/ysar for ths 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, or an incratss in CDP of 3.6E-6. Th3 SRA d:ttrmined thit the loss of redundancy for cooling tha CCPs contributed to increased risk for both a reactor coolant pump seal failure (initiating event) and loss of high pressure safety injection for 'small break LOCAs (mitigating system failure).
The inspectors and SRA screened this finding using the SDP. During the Phase 1 screening the inspectors determined that a Phase 2 screening was required for this issue because it could affect the reactor coolant system barrier function. In conjunction with the SRA, the inspectors initiated a phase 2 screening. During this screening the SRA noted that the CDP for loss of CCW for accident scenarios other than loss of the charging pumps was not affected by unavailability of the SWPs.
SDP Step 2.2 discusses use of Table 1 for findings which relate to an increased likelihood of a specific initiating event and notes that the sample frequency of initiating events should be changed accordingly. For this issue the SRA determined that even though there was an increase in the frequency for an initiating event, the specific CDF calculations indicated that Table 1 should be used without change. The inspectors determined that the " Estimated Likelihood Rating," was an F based on a frequency of initiating event of 1 per 100 to 1000 and a " Exposure Time for Degraded Conditions," of
)
less than three days. Referring to SDP Table 2, the inspectors determined that the j
mitigation capability was two rodundant trains. Although the CCPs were not available, both trains of medium head safety injection pumps were available, and along with operator action to partially depressurize the system, were credited for mitigation, which for a likelihood of F indicated that the issues was within the licensee's response band (green).
The inspectors discussed the observation that three of four methods of cooling the CCPs had not been available for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with the licensee. The licensee initiated a nonconformance report to investigate the issue, perform a separate risk analysis, and determine potential mitigation.
The licensee determined that maintenance personnel could have restored the CCW heat exchanger within 90 minutes prior to seal damage. Based on recovery of the second CCW train, the licensee calculated that the change in the CDF for both SWPs B and D being unavailable for 2 weeks was negligible.
The inspectors reviewed the licensee's analysis that CCW Train B could have been restored within 90 minutes. The inspectors considered that this analysis was adequate; operator mitigation could be credited for restoration of one CCW train before a seal LOCA would have occurred.
The licensee informed the inspectors that the preliminary root cause for this finding was failure to adequately incorporate the unavailability of SWPs B and D into operating, configuration management, and risk management procedures. Licensee planned actions included:
Updating operating procedures (complete) and configuration management procedure (scheduled to by completed within a month of
);
Briefing all operating crews on this issue; Expediting returning the unavailable SWP to a functional status, and not allowing optional maintenance to be performed on the CCW system until SWP B or D was available; Issue Date:
DRAFT E2-9 0610*:
Exhibit 2
Conducting probibilistic riik c:s:ssm:nt-related trcining for s:ltettd maintenance and engine: ring personn:1, including training on tha risk significance of s:rvics water; Reviewing the risk scenarios involving service water to identify additional operator actior;s that could reduce significance; and Updating risk management procedures accordingly.
a 1R07 Inservice insoection Qualification of Inspectors a.
Insoection Scope The inspectors reviewed the ultrasonic testing of core shroud vertical welds by contractor nondestructive examination (NDE) personnel.
b.
_Qbservations and Findinas The inspectors identified a noncited violation for use of an unqualified contract inspector during. performance of core shroud inspections. Reinspection by qualified personnel did not identify any weld cracks.
On July 14,1999, the inspectors, incidental to weld inspection review, identified that the licensee did not have qualification documents for a contractor Level 11 NDE inspector performing the core shroud inspections. In addition, the inspectors could find no evidence that the licensee had reviewed the NDE inspector's qualifications.
The inspectors discussed several requirements with the engineering supervisor in charge of the NDE work. ANSI N45.2.6-1978 required that records of personnel qualifications be maintained by the employer. The Dirojac Quality Assurance Manual required the i
designated technical services engineer to obtain qualified contractors for in-service inspections, and to review NDE personnel qualifications before beginning work. Dirojac Quality Assurance Procedure (DQAP) 320-3, "NDE," Revision 14, required completing a certification checklist before beginning work. Dirojac Test Procedure (DTP) 110-6, "Use of Contractors for NDE," Revision 6, Step 5.2, required preparing a qualification review sheet for each contractor NDE employee.
The engineering supervisor was unaware of whether the necessary certification reviews had been performed. Additionallicensee evaluation determined that the certification review, checklist, and qualification review sheet had not been completed for any contract NDE inspectors. The licensee stopped the core shroud inspection and determined that the qualifications for the inspector in question had expired. After completing the required certification reviews, the licensee repeated the core shroud inspections using a different inspector.
No functional problems were identified by the re-inspection, therefore, the issue did not meet the initial SDP screening, and is considered to be green. The. licensee entered the problem into its corrective action system and determined that the inspector in question had not performed any other inspections at the site. In addition, the licensee determined that all other contract NDE personnel had valid qualifications.
TS 6.2.a requires that written procedures be established, implemented, and maintained a
covering activities recommended in Regulatory Guide 1.33, Revision. 2, Appendix A.
0610*: Exhibit 2 E2-10 issue Date: DRAFT
TS 6.2.c applies to Procedura DOAP 20-3, Procedura DTP 110-6, cnd tha DirojIc Quality Assur2nce Manual. Failura to perform tha r: quired quIlification and certification rzvi:ws before beginning core shroud ultrasonic testing is a violation of TS 6.2.a which is being treated as a nor.dted violation, [ Specific guidance for wording of baseline inspection program NCVs is being developed and will be incorporated into this sample cover letter when it is issued) (NCV 999/99007-01).
1R09 Inservice Testing (Open) Unresolved item 998: 999/98015-06: diesel generator cooling water (DGCW) icsues. Two issues had been identified: (1) the DGCW systems for both Units 1 and 2 were in unbalanced flow configurations, such that flow distribution to individual coolers could not be determined with precision; and (2) the licensee's system flow test did not demonstrate whether the DGCW pump could meet the demands of the diesel generator
. heat exchanger and the Unit 1 ECCS pump room coolers.
Regarding item (1), because of silt accumulation both DGCW systems remain in unbalanced flow configurations. Using existing flow and temperature measurements, Engineering was able to demonstrate (using worst-case assumptions) that sufficient flow existed in each system to maintain system operability. However, to improve flow characteristics and the accuracy of flow distribution measurements, the licensee intended to remove sitt accumulation by hydrolazing the Unit 2 DGCW piping during the upcoming Unit 2 outage. After hydrolazing, the licensee planned to retest the flows to the diesel generator and ECCS room coolers to verify sufficient flow. Flow balancing of individual coolers was not currently planned. Although Engineering Calculation XX appeared to demonstrate that adequate flow existed, this item will remain open pending licensee testing, and subsequent completion of an SDP analysis based on actual data to properly characterize the risk associated with this item and any enforcement based on this risk determination.
l Regarding item (2), the licensee had added an additional test configuration to Dirojac Surveillance Test DST 640-8, " Quarterly DGCW to Unit 1 and Unit 2 ECCS Room l
Coolers Flow Test," Revision 2, to verify that the DGCW pump could meet necessary flow i
demands. On May 23,1999, the licensee's completion of the flow test using the revised procedure successfully demonstrated the capability of the DGCW pump in meeting the flow demands specified above. This portion of the unresolved item is closed.
j j
This item will remain open pending resolution of item (1).
1R14 Nonroutine Plant Evolutions
.1 ECCS Trains a.
Inspection Scope During the Un!12 outage, the licensee determined that the line starter for a Train A sump recirculation valve had failed. The inspectors reviewed the circumstances surrounding -
this failure and the availability of alternate mitigation capabilities.
b.
Observations and Findings 1
Bnef Overview The inspectors determined that both trains of the ECCS had been unavailable for 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> because of equipment failure and planned maintenance which was an apparent Issue Date:
DRAFT E2-11 0610*:
Exhibit 2
viol: tion of TS rcquirsments. Ths NRC str.ff detsrmined thit this issue was within tha incrsestd rrgulitory r:sponsa bind bis:d on mtdium end smril bra:k LOCA exposure times and lack of mitigation.
Backaround At Dirojac, many motor-operated valves are controlled by line starters that direct power to the valve actuators in the opening or closing direction. The control power breaker normally remains closed and the line starter opening and closing coils are de-energized.
To prevent power from being directed to the motor actuator in both directions simultaneously, the line starters are equipped with redundant mechanical and electrical interlocks.
On February 5,1999, while replacing an old line starter for the Unit 2, Train A containment emergency sump outlet valve (2 VAL 84), the licensee discovered that the mechanical interlock was stuck in the closed position. It should have been in the neutral position. In the closed position, the interlock would have preventeo' the valve from electrically opening. The valve had last been operated (closed) on January 6,1999. The licensee documented the deficiency in Nonconformance Report 990500. Unit 2 was shutdown for an outage on January 24,1999.
Problem Assessment On February 6,1999, the licensee determined.that an abrasive foreign material (grit) had worked its way into the space between a metal post and a sliding nylon nng, causing the ring to stick. The licensee also confirmed that the sticking was not beir g caused by another known failure mechanism (a generic problem from 1998, documented in Root Cause Evaluation 98-18, that was the reason that all of that style line starters were being replaced). A formal root cause investigation was initiated.
The licensee established that the grit was not present on any of the newer line starters.
The licensee then inspected all the remaining old line starters in Units 1 and 2 and determined that all the starters were in the required neutral position. Based on this inspection, the licensee considered that.the remaining old line starters were operable (available). The licensee accelerated replacement of the old line starters.
The licensee reviewed the past history of the availability for the line starters, and did not identify any failures related to grit. The inspectors reviewed the maintenance history for selected line starters and did not identify any previous failures. The inspectors considered the failure to be isolated.
Risk Determination The inspectors reviewed this issue with the assistance of an NRC SRA. The stuck interlock mechanism on the Valve 2 VAL 84 line starter would have prevented the valve from opening on demand during a LOCA upon receipt of a recirculation actuation signal (RAS). This would prevent Train A of the ECCS from functioning. The licensee's IPE indicated that RAS was required to prevent core damage in all but the smallest (<
3/8-inch diameter) LOCAs. Without successful initiation of RAS, the licensee determined that core uncovery would occur in 15-30 minutes, and that core damage would occur in another 15-30 minutes.
During the time that the Train A ECCS was inoperable, the Train B ECCS was also inoperable for planned activities on two occasions, as described below.
0610*: Exhibit 2 E2-12 issue Date: DRAFT l
CCW Train B hast exchtngsr was inoperzbis for cpproximitaly 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> for repairs to leaking tubes on January 11-14,1999. Ths lic:n:co estimit:d thit recovery time from this condition would be 2-4 hours, inc'uding reinstallation of the mechanical components and venting the system. The CCW heat exchanger was required to cool the high pressure safety injection (HPSI) pumps and motors. The licensee determined, through its pump vendor, that the pump was designed to operate for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> under accident conditions without cooling water. After that, HPSI unavailability would prevent the successful function of RAS on Train B.
The Train B refueling water storage tank (RWST) outlet valve (2 VAL 123) was inoperable for preventive and corrective maintenance on January 23,1999, for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. The work involved troubleshooting and repairing a motor operator problem. Several components within the operator were being replaced.
Using the licensee's IPE, the NRC staff determined that the incremental increase in risk j
for the combinations of equipment that were unavailable was 3.47E-4/ year for the 18-day
)
period that Valve 2 VAL 84 was unavailable in Mode 1. The CDF for the 18-day period without Valve 2 VAL 84 unavailability was 2.65E-5. Converting these numbers to CDP by multiplying the increased risk for the 18 days (3.47E-4/ year) by (18 days /365 days per year) indicates a 1.7E-5 increase in CDP.
Potential Mitlaation Strateales I
The inspectors discussed various actions that the licensee might have taken in the event of a LOCA to recover ECCS functionality. Much of the information provided by the licensee was preliminary and hypothetical, especially the timing and probability of -
successful performance of various actions, most of which were not contained in procedures.
The licensee determined that the timing of RAS initiation, core uncovery, and core damage were dependent upon the break size. Also dependent upon break size was the time margin provided by the water volume remaining in the RWST following the RAS, which occurs at 18.5 percent level in the RWST. The licensee estimated that operators would continue to run two HPSI pumps and one containment spray pump until the RWST level decreased to 5 percent. The licensee estimated the following times for small break (2-inch), medium break (3-inch), and worst case large break LOCAs:
Small Medium Large Break Break Break RAS initiation in 120 100 47 minutes Useable RWST Time after break / time 140/20 117/17 60/13 volume in minutes after RAS Core uncovery in 260 155/55 78 /31 minutes
/140 The licensee estimated the time it would take to dispatch a team to troubleshoot and effect a temporary repair for the breaker problem. The failure would be revealed at the time of RAS initiation by receipt of a "EMER SUMP VAL 2 VAL 84" annunciator on the
{
i engineered safety featuren bypass status panel, a " RAS INOP" annunciator, an audible Issue Date: DRAFT E2-13 0610*:
Exhibit 2
elirm, end e lo:s of v:lvs position indication for Velve 2 VAL 84. Tha lic:nsu estimitad that a rip;ir trim could b3 dispatch:d within 10 minut:s, end that the 10:m would first identify and replace blown control power fuses. The fuses would blow again when operators attempted to open the valve again. Because of the licensee's recent experience with line starter interlock problems, and the involvement of most of the electricians on site in replacing line starters, the licensee determined that it was likely that the electricians would suspect a faulty line starter. The electricians would then disconnect the coils from the line starter, which does not require determinating the various leads to the coil. This would expose a plastic plunger (part of the mechanical J
interlock) that the electrician would break off, disabling the mechanical interlock. The electrical interlock would still have functioned if needed. The licensee estimated that the breaker could have been returned to service within one hour.
The licensee's maintenance activity on Train B CCW Heat Exchanger 2HX2 involved only the tube (service water cooling system) side. Operators could have run the Train B CCW system, providing cooling to the HPSI pumps and motors. Although the CCW system was closed, and service water cooling would not have been available for heat removal, enough volume was in the CCW system to serve as a significant heat sink for the heat from the HPSI pumps and motors for a substantial period of time. This may have
)
provided enough time, before the core was damaged, for restoration of the service water cooling side of the heat exchanger or repairing the Valve 2 VAL 84 line starter.
}
The licensee's maintenance activity on the Train B RWST outlet valve included partial" motor operator disassembly. The licensee estimated that Maintenance personnel could have reassembled the operator to support opening the valve within two hours.
In each of these strategies, the inspectors considered that a significant variable was the
)
decision-making time. Another significant factor affecting the decisions would be the technical knowledge of the operators and emergency response personnel, particularly
]
regarding the design basis and how much beyond the design basis certain components could function. For example, the licensee could have to determine the minimum RWST level required to provide adequate riet positive suction head to the running ECCS pumps.
Very little time would be available to determine these answers, and a great deal of reliance would have to be placed on the existing knowledge of the licensee's staff.
The licensee performed a preliminary evaluation of the mitigation strategies, including a probabilistic risk assessment that considered those strategies. Some of the assumptions regarding operator actions were validated by running simulator scenarios with two crews.
The licensee concluded that 96 percent of the time at least one recovery action would be successful prior to core uncovery for small break LOCAs.
Sianificance Determination Process The inspectors and an NRC SRA reviewed the licensee's preliminary evaluation and considered it reasonable. The SRA determined that a Phase 2 screening of this finding was required for small, medium, and large break LOCAs. Using SDP Table 1 example frequencies and an exposure time of 3-30 days, the inspectors determined that the
" Estimated Likelihood Rating" was G for a large break LOCA, F for a medium break LOCA, and E for a small break LOCA.
The SRA and the inspectors reviewed the licensee's preliminary mitigation information and considered that the information was acceptable to demonstrate that recovery of equipment for a small break LOCA would have likely been achieved. However, the inspectors considered that the licensee was not able to demonstrate mitigation before core uncovery for medium and large break LOCAs. Therefore, referring to SDP Table 2, 0610*: Exhibit 2 E2-14 issue Date: DRAFT
tha SRA cnd th3 inspectors d:t2rmined th1t thrra would have be:n no mitigition for medium cnd larga br:2k LOCAs end r cov:ry of a fail:d train for a crnill br:sk LOCA.
Table 2 of the SDP indicates for a G (large break) that with no mitigation capability the result is within the licensee response band (green). Table 2 of the SDP indicates for an F (medium break) that with no mitigation capability the result is within the increase regulatory response band (white). Table 2 of the SDP indicates that for an E (small break) that with recovery of a failed train the result is within the increased regulatory response band (white).
The inspectors and SRA discussed their findings and the SDP determination with the licensee. The inspectors asked the licensee to provide the following information; significance attributed to this issue by the licensee, corrective actions and root causes, and position on whether or not an NRC requirement had been violated. The licensee stated that it would provide the requested information.
Reauirements TS 3.5.2 requires that two trains of the ECCS be operable in Modes 1,2, and 3, with pressurizer pressure greater than or equal to 400 psia. Valve 2 VAL 84 not being capable of opening during a recirculation actuation rendered Train A of the ECCS inoperable from January 5-24,1999. Train B of ECCS was concurrently inoperable on January 11-14 and January 23,1999.
Pending receipt of the information requested from the licensee, this is an apparent violation of TS 3.5.2 (eel 50-999/99007-02).
Corrective Actions to Date The licensee replaced all the remaining old line starters in Unit 2 during the latest outage and began replacing the remaining old Unit 1 line starters on line. All the line starters have been replaced.
The licensee initiated a root cause analysis. The licensee has been unable to determine the source of the grit as of the end of this inspection period.
Summary The inspectors identified an apparent violation of TS Limiting Conditions for Operation 3.5.2 in Unit 2 as a result of both trains of the ECCS being unavailable for approximately 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> and Train A being unavailable for approximately 18 days during power operations. The simultaneous unavailability of both trains resulted in the totalloss of a mitigation function necessary to prevent core damage in the event of a LOCA. The Train B unavailability was planned. However, the Train A unavaitability, caused by mechanical failure of a line starter providing power for a Train A containment emergency sump outlet valve, occurred before the Train B unavailability and was not discovered until later. During that time, Train A would not have functioned following a recirculstion actuation signal. Staff calculations indicated an yearly increase of 1.7E-5 in core damage probability. Based on Tables 1 and 2 of the SDP the NRC staff determined that the screening for a small break LOCA was within the increased regulatory response band (white); low frequency and medium likelihood (E) with recovery of one train. The screening for a medium break LOCA was also within the increased regulatory response band; lower frequency and medium likelihood (F) with no mitigation Issue Date:
DRAFT E2-15 0610*:
Exhibit 2
.2 (Closed) Licensee Event Reoort (LER) 999/1998-004-00: subcritical rcictor trip b causs of inadequata procedura. Following a shift chings, oncoming control room operators did not initiate auxiliary feedwater soon enough during preparations for Unit 2 startup. A procedure interface problem was identified, involving Dirojac Operating Procedure (DOP) 143, " Unit Startup," Revision 2, and DOP 512, " Steam Generator Crevice Flushing,"
Revision 4. DOP 512 referred the operators back to the wrong portion of DOP 143, thus omitting the step requiring initiation of auxiliary feedwater.
The licensee promptly corrected this procedural inadequacy. The inspectors reviewed
. this LER and determined that the trip was uncomplicated, all mitigation systems were available, and reactor coolant system barrier integrity was not challenged. Even though the reactor was not at full power, the inspectors compared the event to the reactor SDP initial screening criteria. The SDP Phase 1 criteria includes a statement that if only the initiating event comerstone is affected and associated assumptions have no other impact than increasing the likelihood of an uncomplicated reactor trip, the finding would be considered green and screened out. Therefore, the inspectors considered the event was 1
within the licensee response band (green).
The inspectors determined that, at the time of the event, DOP 512 was not appropriate to the circumstances, constituting a violation of 10 CFR Part 50, Appendix B, Criterion V,
" Procedures." This procedure violation is being treated as a noncited violation [ Specific guidance for wording of baseline inspection program NCVs is being developed and will be incorpomted into this sample cover letter when it is issued] (NCV 999/99007-03).
1.R17 Permanent Plant Modifications a.
Inspection Scone The inspectors reviewed Dirojac Design Modification CCW96-1, " Addition of instrument Wells," Revision 0, in use during power operations.
b.
Observations and Findinos There were no findings identified and documented during this inspection.
1R19 Postmaintenance Testina
.1 Untimelv Retest Followina Scram Inlet Valve Packina Adiustment a.
Insoection Scope The inspectors reviewed the PMT for Dirojac Maintenance Procedure M51, " Adjustment of Scram inlet Valve Packing," Revision 0, under Work Package 1-96MW3117 l
b.
Observations and Findinas -
The inspectors determined that a PMT was not performed prior to retuming a control rod to service. The licensee subsequently performed the PMT.
in attempting to witness performance of the specified retest on June 18, the inspector l
found that Work Package 1-96MW3117 was still filed as not completed, because Control j
Rod J-10 had not yet been tested for scram time in accordance with the retest portion of i
the package. However, in a tour of the control room, the inspectors found that Control Rod J-10 was in service. A review of the June 13,1999, operator logs revealed that the 0610*: Exhibit 2 E2-16 issue Date: DRAFT
e control rod had baan r tumtd to opgrational status following tha ceram inl t valva packing adjustment.
The inspectors brought this matter to the attention of the shift engineer. After reviewing the work package, with Unit 1 operating at full power, the shift engineer declared Control Rod J-10 inoperable. Operators fully inserted the rod, then successfully completed scram time testing. The work package was then filed as complete, and the control rod returned to service.
Later discussions confirmed that the control rod had been inserted and removed from service during the packing adjustment. At the completion of the adjustment, the shift engineer had asked the maintenance foreman for the retest requirements. Instead of reviewing the retest ponion of the work package, the maintenance foreman had reviewed a vendor memorandum and incorrectly determined that a scram time test was not required. The shift engineer had not questioned the foreman's assessment, and had not independently verified the postmaintenance testing requirements. The control rod was retested without adjustment. The licensee entered this problem in their corrective action system. The inspectors determined that this error did not lead to any measurable change in plant risk; therefore, the issue did not meet the initial SDP screening, ard is green.
10 CFR 50.55a and Section IWV-3200 of ASME Section XI (1986) require that after valve stem packing is adjusted, and before it is retumed to service, the valve shall be tested to demonstrate that performance parameters are within acceptable limits. Failure to.
perform a scram time test after tightening the packing and before retuming Control Rod J-10 to service on June 13,1999 is a violation of these requirements. This procedure violation is being treated as a noncited violation, [ Specific guidance for wording of
)
baseline inspection program NCVs is being developed and will be incorporated into this sample cover letter when it is issued) (NCV 998/99007-04).
~
.2 (Closed) Violation 998/96003-03: failure to provide adequate procedural guidance for check-valve inspection retest. Dirojac Surveillance Procedure 740-2, "HPCI Torus Suction Check Valve Inspection," Revision 1, did not verify that the disc would properly seat after the check valve intemals were reassembled. The licensee had revised the procedure to incorporate a seat tightness test (using a feeler gauge) after valve intemals were reassembled in addition, for check valve disassembly, the licensee completed a data sheet to document the type of seat leakage check to be performed on each valve. The inspectors reviewed three subsequent work requests involving check valve disassembly and inspection, and confirmed that a seat leakage test had been properly documented in each case.
2.
RADIATION SAFETY 2OS2 ALARA Plannina and Controis a.
Insoection Scope The inspectors reviewed ALARA planning for the radiological controls implemented in the Unit 2 refueling outage, b.
Observations and Findinas Outage dose to date was about 315 rem with about 70 percent of the work completed.
The accumulated exposure was on target for the licensee to meet the planned exposure goal.
Issue Date:
DRAFT E2-17 0610*:' Exhibit 2 i
L 1
2OS4 Radiation Worker Performance -
a.
insoection Scope The inspectors observed an RP technician on a tour of the Unit 1 radiological protected area (RPA) and outside areas to observe and discuss radiological control practices.
b.
Observations and Findinas RP technicians failed to remove all of the tools and other material with low levels of radioactive contamination prior to release of a trailer from use as a temporary RPA. The licensee had recently identified two similar release problems on radiological problem j
reports.
During the tour of areas outside the RPA (but inside the restricted area), the RP technician discovered a yellow bag of contaminated material and several contaminated tools inside the motor-operated valve trailer, a recently released RPA. The bag was labeled as containing material with contamination levels of 3,000 - 115,000 dpm. The tools were
, painted purple, denoting fixed contamination. The RP technician took prompt action to secure the trailer, perform additional surveys, and post the area.
Subsequent licensee investigation found that the trailer had previously been posted as an i
RPA and radioactive materials storage area (RMSA), but had been released from RPA and RMSA status on June 30. The trailer had been surveyed and released by a contractor RP technician. The Unit 1 RP manager informed the inspectors that licensee policy did not normally allow contract technicians to release an area from RPA status.
The licensee performed additional surveys that demonstrated that the radioactive material contained in the yellow bag was less than 10 times the quantity of licensed material j
specified in Appendix C to 10 CFR 20.1001 - 20.2401. The inspectors concluded that the l
lack of radiological posting on the motor-operated valve trailer did not constitute a violation; however, the inspectors' review of radiological problem reports disclosed two other recert instances in which the release of temporary RPAs had not been well 4
controlled. The Unit 1 RP manager stated that the effectiveness of RPA release policies would be revewed. The licensee had entered all of the problems discussed above in their corrective actonti system.
)
4 OTHER ACTIVITIES 40A4 Other (Closed) LER 998/1998-001-00: auxiliary building ventilation actuation. This LER was a minor issue and was closed.
40A5 Manaaement Meetinas
.1 Exit Meetina Summarv i
The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 24,1999. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection j
should be considered proprietary. No proprietary information was identified.
0610*: Exhibit 2 E2-18 issue Date: DRAFT
.2 Predecisional Enforcement Conference Summary On July 24, a predecisional enforcement conference was held at the NRC Region I office to discuss potential enforcement issues identified in NRC Inspection Report 50-998i 999/99-06. The issues related to radiological concerns over the licensee's control of access to high radiation areas. Slides used in the licensee's presentation at the conference have been included as Attachment A to this report.
1
)
J Issue Date: DRAFT E2-19 0610*:
Exhibit 2
PARTIAL LIST OF PERSONS CONTACTED Licensee J. Cramer, Outage Supervisor
' J. Delphi, System Engineering Supervisor '
G. Deplogie, Maintenance Manager, Unit 1 S. Nithhold, Manager, Quality Assurance G. Picket, Radiation Protection Manager, Unit 2 D. Prue, Operations Manager, Unit 2 J. Russelville, Radiation Protection Manager, Unit 1 J. Sloaninton, Manager, Compliance L. Smithson, General Manager, Technical Services J. Spots, Supervisor, Mechanical Maintenance, Unit 1
)
NBC D. Ackerman, Dirojac Project Manager lTEMS OPENED, CLOSED, AND DISCUSSED Opened 999/99007-02 eel Apparent violation of TS 3.5.2 for ECCS operability Z
Opened and Closed 999/99007-01 NCV failure to review NDE inspector qualification 999/99007-03
. NCV reactor trip because of procedure problem 998/99007 NCV failure to retest control rod after maintenance Closed 998/98003-03
_VIO failure to provide adequate procedural guidance for check valve inspection retest 998/1998-001-00 LER missed surveillance because of personnel error 999/1998-004-00 LER reactor trip while suberitical because of inadequate procedure Discussed i
998/98015-06 URI diesel generator cooling water issues j
LIST OF BASELINE INSPECTIONS PERFORMED The following inspectable area procedures were used to perform inspections during
\\
the report period. Documented findings are contained in the body of the report.
Emergent Work R03 Equipment Alignments R04 Heat Sink Performance R07 Inservice inspection R08
' Inservice Testing R09 Large Containment Veb s R10 0610*: Exhibit 2 E2-20 Issue Date: DRAFT
Maint:ntnce Rul2 ImpirmIntation R12 Miint:n nce Work Prioritization R13 NonRoutine Events R14 Operability Evaluations R15 Permanent Plant Modificationsw R17 Postmaintenance Testing
~
R19 Surveillance Testing R22 Radioactive Material Shipping -
PS2 Radiological Environmental Monitoring PS3 Access Control OS1 ALARA Planning and Controls OS2 Radiation Worker Performance OS4 PI Verification OA2 Meetings, including Exit OAS LIST OF ACRONYMS USED ALARA as low as reasonably achievable CCP coolant charging pump CCW component cooling water CDP core damage probability CFR Code of Federal Regulations DGCW diesel generator cooling water DMP Dirojac Maintenance Procedure DOP Dirojac Operation Procedure DOAP Dirojac Quality Assurance Procedure ECCS emergency core cooling system eel escalated enforcement item GP&L Greckenshire Power and Light HPCI high pressure core injection j
HPSI high pressure safety injection IPE individual plant evaluation LER licensee event report LOCA loss of coolant accident NOV noncited violation NDE nondestructive examination NRC Nuclear Regulatory Commission PMT
.postmaintenance test RAS recirculation actuation signal RMSA radioactive materials storage area RP radiation protection RPA radiologically protected area i
RWST refueling water storage tank SDP significance determination process
-SRA.
senior reactor analyst S W P.-
service water pump 1
TS technical specification
-1 Issue Date:
DRAFT E2-21 0610*:
Exhibit 2
NONCOMPLIANCE INFORMATION CHECKLIST Anoendix A NOTE: This checklist is presented as a guideline for gathering and arranging enforcement-related information.
However, it should not be considered prescriptive, nor in all cases will it be all-encompassing. Therefore, use this list
. only as a reminder of items to consider; determining the answers to all or many of these questions will often be beyond what is required to adequately describe the violation in the baseline inspection process REQUIREMENT -
O What requirement or commitment was violated? If the requirement was conditional, how were the conditions satisfied which made the requirement applicable?
STATEMENT OF VIOLATION O
How was the requirement or commitment violated?
O By whom (individuals and tities) was the requirement or commitment violated?
O When was the requirement or commitment violated and what was the duration of the violation?
CONTEXT O
What were the circumstances surrounding the violation (such as system configuration and operational conditions for reactor cases)?
O How, when, and by whom (licensee or NRC) was the violation discovered?
O When was the licensee aware or put on notice of the problem requiring corrective action?
O Was the violation the result of a self-disclosing event?
ROOT CAUSE/ CORRECTIVE ACTION O
What was the apparent root cause (and contributing causal factors) for the violation?
O What short-term corrective and remedial action was taken and when was it taken?
O-Did NRC have to intervene to accomplish satisfactory short-term corrective and remedial action and, if so, to what degree?
O Were the licensee's corrective actions comprehensively or narrowly focused?
ENFORCEMENT SIGNIFICANCE O
What was the actual safety consequence of the violation (e.g., overexposure, release of
- radiation, loss of redundancy, inoperable safety system, degraded system, etc.)?
O What was the potential safety consequence of the violation?
O Are there other circumstances surrounding the violation which increase or decrease its significance (e.g., appearances of willfulness, careless disregard)?
NOTE: Inspection documentation should describe the circumstances of the violation in a j
manner to support later discussions regarding enforcement action. However, inspection reports should NOT offer conclusions regarding safety significance, willf ulness, or careless disregard, as these are agency conclusions.
O Was management aware or should it have been aware of the violation?
)
O ls there evidence that management was involved directly or indirectly in the violation and j
to what extent?
Issue Date:
DRAFT A-1 0610*: Appendix A
O ls the violition a r pItitiva vioittion or similrr to past violations? If co, should th3 previous correctiva actions h;va be:n cdequit3 to pr;v:nt recurrence?
O Are the inspection findings of programmatic problems?
ADDITIONAL FACTORS O
If the violation was a result of a self-disclosing event, did the licensee demonstrate initiative in identifying the root cause?
O Were there prior opportunities for the licensee to identify the violation, such as through audits or NRC or industry notification that would have reasonably put the licensee on notice of the potential for a violation?
O-Do the inspection findings represent another example of poor performance or do they represent an isolated occurrence?
O Were there multiple examples of a particular violation?
O Did the duration of the violation add particular significance to the issue?
REPORTABILITY O
If the violation or the conditions leading to the violation were required to be recorded and the matter was not properly recorded, what was the applicable recording requirement?
O Was the violation required to be reported and, if so, what was the applicable reporting requirement?
O Was the violation reported and, if so, when and by whom was it reported?
_ f the violation was reported, but the report was late, why was the report late?
l O
O Was the report complete and accurate?
l 0610*: Appendix A A-2 Issue Date:
DRAFT 1
MANAGING THE WRITING PROCESS Aooendix B a.
==
Introduction:==
Writina Efficientiv and Cleariv. Few inspectors place report writing high on their list of favorite activities. For many, the writing process is a frustrating, laborious experience invoMng many revisions. Contrary to popular belief, however, techniques can be leamed that will make the writing experience more manageable.
Leaming to present facts in a clearly developed, logical order is a key ingredient in developing your communication skills. By consistent practice, you will recognize patterns of thought, repetitive arrangements of similar facts, that will increase both the speed and effectiveness with which you write. In addition, these familiar, repetitive ways of thinking will help you ask key questions earlier in the inspection process, thus improving your field inspection skills as well as your writing.
b.
Seoaratina the Parts of the Writina Process. Writing involves planning, brainstorming, organizing, drafting, and revising. Trying to perform all of these activities at once will produce " writer's block," incoherent presentation of findings, sloppy final drafts, and extended cycles of report review and revision. A more effective writer breaks the writing process into stages to avoid frustration, save time, and produce a more professional product.
I 1.
Plannina. For the inspector, planning is the simplest stage of the report writing i
process; however, it asks two important questions that will help to maintain later focus:
What is the purpose and scope of this inspection report?
Who are my readers, and what are their priorities?
Ideally, these questions will be answered before beginning the inspection. The purpose of the inspection report will vary slightly with the nature of the inspection findings (e.g., depending on whether or not significant weaknesses or enforcement matters were identified). The scope of the inspection report will be similar to the scope of the inspection itself, which in turn may be largely determined by the scope of a well-prepared inspection plan.
The report's principal reader, as stated earlier, will be the vice president, nuclear (for a reactor facility), or a similar official.
This reader's priorities will include understanding your findings (and perhaps defending a differing perspective), evaluat-ing your conclusions, improving facility safety, protecting the environment, continuing unhampefed operation, and maintaining good public relations. Being aware of these reader priorities will not change any of the inspection results (i.e., the report observa-j tions, findings, and conclusions); however, it will help to make you a'better communicator (e.g., in setting the proper tone for the cover letter, or in emphasizing the need for written conclusions to be well supported by facts).
In addition, never expect to become a skilled report writer unless the inspection itself is well-planned.
The importance of a detailed inspection plan cannot be overemphasized; it will not only influence the effectiveness of your inspection, but will also directly affect the amount of time you spend in writing. A good inspection plan is your thst report outline.
2.
Brainstormina. Brainstorming involves the rapid, noncritical dumping of ideas onto paper (or into a computer). Like planning, brainstorming will first occur during the construction of an effective inspection plan. In addition, you will brainstorm during Issue Date:
DRAFT B-1 0610*:
Appendix B
I' glmo:t rony insptction, as you specullt3 on the implications of your obs:rvations and in;pect ntw gr:20.
In writing reports, brainstorming is the single most effective weapon against writer's I
block. When presenting a difficult technical issue, writer's block may result from the unrealistic ambition of trying, simultaneously, to remember all the necessary details, balance each one against the appropriate regulatory criteria, analyze the major areas of discrepancy, and blurt out the whole conglomeration as a coherent, well-organized report section.
Instead, brainstorm by listing all the related ideas rapidly on paper or computer, working from memory and your inspection notes. Avoid worrying about organization or relative importance or what the report reviewer will think. Your brainstorming is l
complete only when you have assured yourself that all details relevant to this report l
section have been written down. At that point, you are ready to consider the next l
step of the writing process.
l l
3.
Oroanizino. Organizing naturally follows brainstorming. Group your brainstormed ideas into topical categories, then arrange each the ideas in each category in a logical order. If the scope of your inspection did not change significantly, these i
categories and their order may correspond naturally to the outline already provided l
by your original inspection plan. Standard methods of organizing details include:
(
Order of linoortance. Within each category, present the observations and j
findings beginning with the most significant.
Chronoloaical. This method is useful in presenting the essential details of.a L
complicated event.
Comoarison/ Contrast. This is the organizational method used for writing l
Notices of Violation, in which the regulation is quoted or paraphrased and closely parallel language is used in the " Contrary to" statement. This is also an effective method of presenting a negative finding in the report details.
In addition, you may find it helpful to construct " repetitive formats" as organizational models for certain types of findings where the underlying structure is generally the L
. same from one report to another (such as when documenting events, follow-up of open items, noncompliances, review of audits, etc.). This technique is an extremely effective time-saver. Section 07.02.c of IMC 0610 gives additional guidance on how l
to create and use these " repetitive formats."
4.
Draftina. Now you are ready to draft the report. Most of the distaste inspectors feel for report writing is focused on this stage of the process. However, as stated earlier, frustration in writing is often the result of trying to do too many things at once. By breaking the writing task into separate, manageable steps, you make drafting itself l
a shorter, less complicated process.
If you have planned, brainstormed, and organized well, and if you resist the i
temptation to revise as you go, you can write your draft rapidly and effectively. This usually involves unleaming bad habits, and requires practice. Follow these drafting steps:
(a)
Work from your outline of categories and brainstormed, organized notes.
(b)
Begin with any section.
i 0610*: Appendix B B-2 Issue Date:
DRAFT l
L
(c)
Writs quickly!
(d)
Resist the temptation to edit!
Step (d) is, by far, the most difficult. Most writers have leamed to re-evaluate spelling, syntax, organization, and even the value of the written content while they are actually drafting. By resisting this acquired habit, the drafting process speeds up, revision becomes a natural part of the writing process, and the result is both more polished and more quickly produced.
5.
Revising. Always let your draft " cool off" before beginning to revise. If you revise while you draft, or revise immediately after drafting, you will tend to see the inspection report in your mind's eye rather than the one on paper. As a result, you will be vulnerable to missing obvious mistakes.
In addition, never submit this first draft to your supervisor for review. Supervisors often take on the role of technical editor, making the review and concurrence process into a frustrating cycle of revised revisions. You can avoid this frustration by leaming to responsibly edit your own work.
Levels of revision should be prioritized. Some corrections are more important than others; if you begin by focusing on spelling and punctuation, you may be too distracted to notice larger organizational problems. The most effective practice is to revise in several stages.
The first stage is a rapid read-through to check for overall coherent organization and level of detail in each report section. if no major revisions are needed, the second
+
stage review is then performed, reading to assess paragraph structure, sentence style, and clarity of syntax. The final stage is a careful word-by-word proofread to check for spelling, punctuation, and accuracy. An " Inspection Report Review Checklist" that explains each of these revision stages in detail is given in IMC 0610, Appendix C.
c.
Understandina the Suoervisorv Review Process The inspection report review and concurrence process is frequently an area of inefficiency.
For both supervisors and inspectors, it can become the focus of considerable frustration.
Much of this frustration comes from miscommunication, unclear standards, and misunderstood roles. By understanding the review process and applying intelligent effort, you can help to make these interactions positive and productive.
1.
T.h.e importance of Suoervisorv Review. The inspector and the supervisor are in some ways like liaisons to the rest of the world. The inspector, as the NRC's " eyes and ears," interacts with the licensee or vendor with a responsibility to observe accurately, analyze logically, and report the results. The supervisor's responsibility is to ensure that when these results retum to the "outside world" in their final written i
I form, they reflect past enforcement practice, current industry knowledge, and approved agency emphasis, in producing high-quality inspection reports, therefore, the supervisor serves a vital (but often misunderstood) role. The supervisor is a shield between the inspector and certain " hazards,* such as unintentional backfits, inconsistent agency positions, and retracted violations. In the role of liaison, the supervisor anticipates the reactions of the report's readers (such as the licensee, NRC headquarters and regional management, the Office of Enforcement, and the public). The supervisor also Issue Date:
DRAFT B-3 0610*: Appendix B
~
gnsurzs that th3 r: port's missaga is ci:sr and appropriita in t rms of safaty significance, indu try precedInt, agsncy position, and g:n:ral tonn.
Conversely, the supervisor should not be viewed as a technical editor. The
-supervisors function should not be_ to correct spelling errors, break up long paragraphs, and fix problems of report organization. In such cases, he or she can easily become so immersed in the report details that the more important supervisory functions described above are lost.
When the supervisors role becomes blurred in this manner, a cycle of inefficient review and concurrence begins. The draft report circles in a painful path from the supervisor to the inspector to word processing and back again. By understanding tM proper importance of supervisory review, and by ensuring that you make thorough, accurate revision your own responsibility, you can avoid this inefficiency and save yourself frustration.
2.
Clarifvina the Suoervisor's Writina Expectations. Despite the best efforts of English instructors, writing remains a changing and subjective science. IMC 0610 and other r
references (the Fundamentals of Inspection Course, the NRC Enforcement Manual.
etc.) establish a standard style and format for inspection reports; some aspects of the writing, however, remain a matter of personal preference. Your supervisor may have specific additional standards in these areas.
When the review and concurrence process has become the focus of frustration, it may be because of ill-defined supervisory expectations. If the supervisor's writing standards differ from your own, you will both save a great deal of effort by taking the time to clarify those standards.
(a)
Ask for a "model report" that meets your supervisors standards. If possible, ask your supervisor to make clarifying remarks in the margins of the model report about what he or she expects.
(b)
When changes are made to your writing, try to decipher' the rule behind the change. Ask for clarification that will make your supervisor's comments useful for future reports (e.g., "Do you prefer that I always refer to the regulatory criteria before discussing the related findings?").
3.
Coooerative Writer / Reviewer Relationshio. A cooperatrve inspector / supervisor relationehip must be leamed, in many cases. Old bad habits don't just die hard-they also give birth to new bad habits. For example, a non-cooperative writer / reviewer relationship may condition you to expect that your reports will always require significant revision; as a result, you may become complacent about submitting a report that you know contains errors.
Clarifying the supervisors writing standards, as discussed above, is one step in changing the "old habits." Three more are discussed below: early supervisory i
involvement, effective debriefings, and submitting final drafts.
(a)
Early Suoervisory involvement. If you wait to involve your supervisor in the inspection findings until atter the inspection, you should expect to have your findings challenged at the debriefing. Worse still, if you first tell your supervi-sor about the findings by handing in a draft report, the review and concurrence process may become lengthy simply in order to give the supervisor time to mull over your ideas.
l 0610*: Appendix B B-4 Issue Date:
DRAFT o
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' Supervisory ravi;w," id: lly, thould begin et tha inspection plinning stig).
A short bri;fing, cov: ring your int nd:d inspection scope, will piant ths er:12 of focus in your supervisor's mind, begin the cooperative effort, and build respect.
If your supervisor tends to respond to your findings by suggesting additional
)
areas for inspection, you will profit by getting that input as early as possible.
Telephone discussions with your supervisor several times during the inspection will continue the cooperative effort (brainstorming is even more effective with two people).
(b)
Effective Debriefin_22 As a matter of routine, hold a short telephone discussion with your supervisor before the exit meeting. By evaluating the inspection findings at this stage, you will avoid inconsistencies between the exit meeting and final report.
In any event, be sure to hold a thorough, organized debriefing with your 1
I supervisor before beginning the report. Two rules apply:
(1)
Always try to resolve differences of opinion before ~ writing about them.
(Sometimes you will hear a statement like, "Well, why don't you try writing it up and we'll see how it looks on paper...." Using this approach to resolving technical or enforcement issues is a sure avenue j
to inefficiency. Disagreements take much longer to resolve on paper j
than in oral discussion. Written products should become the arena for J
resolving differences only when unavoidable.)
1 (2)
Give the supervisor a mental picture of what the report will look like. By preparing and presenting your findings in a well-organized debriefing, your expectations about the content and tone of the final report will become aligned with your supervisor's.
1 (c)
Submittina Final Drafts. If possible, never give a reviewer anything other than a final draft (i.e., don't hand in a draft before you have given it your own best i
revision efforts). The presence of errors--even minor spelling and punctuation errors--will predispose any reader to believe the document is also flawed in other ways.
l Issue Date: DRAFT B-5 0610*:
Appendix B
INSPECTION REPORT REVIEW CHECKLIST Appendix C The checklist can be an effective aid both for writers and reviewers. It is not intended to be a prescriptive recipe; used consistently, however, it will add considerable focus and efficiency to the writing / reviewing process, and it will improve the clarity of the written product.
i First Staae Review Cover Letter.
l
{
i
-O Major idea is clearly presented O<
Message is supported by findings in the Summary of Findings O
Letter uses appropriate tone
{
S,ymmarv of Findinas O
Organization follows the order of the standardized report outline i
O Each significant finding is given a clear, concise description i
O Summary of findings are supported by report details Reoort Details: Overall Oraanization O
Organization follows the order of the standardized report outline O
Each report section uses standard intemal organization O-
" Areas inspected" sub-sections are presented clearly and concisely Report Details: Presentation of Reoort Observations /Findinas/ Conclusions O
Thresholds of significance (in determining what to document) are appropriate O
Findings are clearly developed Main ideas are clearly presented Observations are placed in context Assertions are supported by facts Requirement or standard is included, where appropriate Licensee response to findings is included, where appropriate O
Level of detail is appropriate (based on significance, complexity, and reader awareness)
Issue Date:
DRAFT C-1 0610*: Appendix C r
Sagond Staae Review
,Qlear oresentational style O
Subheadings are used where appropriate O
Emphasis techniques (bullets, underlining) are used where appropriate O
Statistical data is presented in graphic or tabular format Paracraohs O
Each paragraph develops only one main idea O
Main idea of the paragraph is clearly presented O
Para, graphs are generally a maximum of 10-12 lines Sentences O
Active voice predominates (Subject-verb-object)
O Wordy phrases are avoided (e.g., "Because of the fact that..." vs. "Because...")
O False subjects are avoided
,, (e.g., It is clear that..." vs. " Clearly...")
O Weak verbs are avoided (e.g., "The licensee gave authorization to..." vs. "The licensee authorized...")
O Word selection reflects a generally understood vocabulary (except where the use of technical jargon is necessary)
Third Staae Review 0
Cover letter, cover page, summary of findings, and report details are accurate and consistent on all dates, distribution, titles, etc.
O Acronyms and initialisms are defined O
Grammar and punctuation are correct O
Spelling is correct l
-0610*: Appendix C C-2 Issue Date:
DRAFT
1 LIST OF ACRONYMS USED IN THIS INSPECTION MANUAL CHAPTER Aooendix D NOTE: a separate list of acronyms is given as an enclosure to Exhibit 2, the sample
)
inspection report.
1 AEOD Office for Analysis and Evaluation of Operational Data ALARA as low as is reasonably achievable CFR Code of Federal Regulations CVCS chemical and volume control system EA escalated action EP emergency preparedness ESF engineered safety feature EW exercise weakness gpm gallons per minute GPO Govemment Printing Office IFl inspection follow-up item IFS Inspection Follow-Up System IMC inspection manual chapter IPAP Integrated Performance Assessment Process IRAM ltem Reporting and Analysis Module ISI in-service inspection LER licensee event report MD management directive NCV noncited violation NMSS Office of Nuclear Material Safety and Safeguards NOV notice of violation NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation OE Office of Enforcement 01 Office of Investigations PIPB Inspection Program Branch PPR plant performance review RA regional administrator RHR residual heat removal RP radiation protection RP&C radiological protection and chemistry SDP Significance Determination Process Si international System of Units Tl temporary instruction TS technical specification l
Issue Date:
DRAFT D-1 0610*: Appendix D 1
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i inspection Manual Chapter 06XX 4
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NRC INSPECTION MANUAL r
PIPB Manual Chanter 06XX SIGNIFICANCE DETERMINATION PROCESS 1
06XX-01 PURPOSE l
To provide guidance for the si nificance determination (Risk Characterization)de
~
in this of an in etion l
program finding. The in finding sig,nificance determination processes procedure and its appen ixes evaluate the significance of individual inspection findings so that the overall licensee performance assessment process can compare and evaluate them on a significance scale similar to the plant performance indicator (PI)for this process.
information. Licensee-identified i
i issues, when reviewed by NRC inspectors, are also candidates 06XX-02 OBJECTIVE 02.01 To characterize the risk significance or importance of an inspection finding consistent with the regulatory response thresholds used for performance indicators (Pis) in the NRC licenseo performance assessment process.
1 1
02.02 To provide a risk-informed framework for discussing and communicating the potential significance of inspection findings.
02.03 To provide a basis for assessment or enforcement actions associated with an inspection finding.
02.04 To specify the minimum amount of documentation needed to allow reconstruction of the basis for any decisions associated with the risk significance ranking of an inspection finding.
06XX-03 DEFINITIONS Anoarent Risk Sianificant issue. An issues that has been processed through the SDP and its risk estimation is greater than that associated with a Green finding.
Findina. As used in this chapter, an observation that has been placed in context and assessed for significance.
Observation. A fact; any detail noted during an inspection.
Sionificance Determination. The process for applylng a risk characterization to an individual issue for the purpose of providing an i0 ggt to the NRC's Reactor Oversight Plant Assessment and l
Enforcement Processes.
06XX-04 RESPONSIBILITIES AND AUTHORITIES All NRC inspectors are required to assess the significance of inspection findings in accordance with i
the guidance provided in this Inspection Manual chapter. General and specific responsibilities are listed below.
i Issue Date: 05/07/99 -
DRAFT 06XX
04.01 Director. Office of Nuclear Reactor Reaulation.
a.
Provide overall program direction for the reactor inspection program.
b.
Develop and direct the implementation of policies, programs, and procedures for regional application of the S nificance Determination Process in the evaluation of findings and issues associated the Reactor Oversight Program.
c.
Assess the effectiveness, uniformity, and completeness of regionalimplementation of the SDP.
. 04.02 Aswinte Director for inspection and Proarams.
Direct the development of the SDP within NRR 04.03 Director. Division of Insoection Proaram Manaoement '
Provide oversight and representatives as necessary to support the Significance Determination Process Oversight Panel in order to ensure consistent application of the process.
04.04 Directors. Technical Division. NRR.
Provide technical oversight and prepare periodic revisions and guidance associated with the implementation and application of the SDP in their areas of responsibility.
04.05 Director. Office of Enforcement a.
Ensure consistent application of the enforcement process to violations of NRC regulations with the appropriate focus on the significance of the issue.
b.
Provide representatives as necessary to support the Significance Determination Process Oversight Panel in order to ensure consistent application of the process.
04.06 Director. Office of Research a.
Provide support in the development and refinement of the SDPs, which use risk insights from research activities.
b.
Provide representatives as necessary to support the Significance Determination Process j
Oversight Panel in order to ensure consistent application of the process.
04.07 Realonal Administrator Provide program direction for mana performed by the Regional Office. gement and implementation of the SDP to activities a.
b.
Provide representatives as necessary to support the Significance Determination Process Oversight Panel in order to ensure consistent application of the process.
c.
Within the guidance of the Reactor Oversight Program, apply inspection resources, as necessary, to determine the significance of specific issues identified.
06XX-05 BASIC REQUIREMENTS INSPECTION FINDING SIGNIFICANCE DETERMINATION PROCESS (SDP) introduction SECY-99-007, dated January 8,1999, described the need for a method of assigning a risk characterization to inspection findings. This risk characterization is necessa so that inspection findings can be aligned with risk-informed plant performance indicators (P during the plant performance assessment process. Figure 1 describes the process flow o ical inspection 06XX DRAFT Issue Date: 05/07/99
l a
i findin or issues. Figura 1 elso outlin:s th3 diffzr:nt p;ths an issus could taks with tha final of each process b:ing an inDg: to tha assessm:nt and/or tha enforc:m:nt process.
J rcement associated with violations o regulatory requirements will be processed in accordance 1
with NUREG-1600, Rev 1, General Statement of Policy and Procedures for NRC Enforcement Actions and any applicable Enforcement Guidance Memorandums (EMGs).
Appendix 1 of this attachment describes the significance determination of inspection findings, which j
have a potential impact on power operations, thereby affecting the initiating event, mitigating systems, or barrier cornerstones associated with the reactor safety strategic performance area.
It is e ed that this process will address most of the risk-significant issues that would be need at a facility. Issues associated with, emergency preparedness, radiation safety, e
oguards, fire protection and shutdown risk also needs a SDP as well. Appendix 2 of this attachment provided the SDP processes for emergency pr,eparedness, radiation safety, and safeguards. Appendices' 3 and 4 associated with findings in the fire protection and shutdown area will be provided later after additional staff development and review.
i Although the staff fully expects to have most of the risk characterization processes in place for the pilot study, further enhancement and development will continue. However, if for example, difficulty is encountered in developing a method for the risk characterization of issues associated with fire protection and shutdown activities, the inspection staff may have to involve a risk anayst or the SDP review panel as described in appendix 5 in order to property characterize the finding until guidance can be developed.
Issue Date: 05/07/99
-3 DRAFT 06XX
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- P FARD XX 60
Appendix 1 Significance Determination of Inspection Findings for Power Situations Entry Conditions i
This process is currently designed to assess only those inspection findings associated with power operations within the comerstones of initiating events, mitigation systems, and barrier integrity under the reactor safety strategic performance area. Compliance with Technical Specifications (TS) and design-bases assumptions continue to provide defense-in-depth and safety margins. This process was developed to provide the risk insight for conditions that may affect the above-mentioned assumptions. An actual initiating event will either be captured by a performance be asses (e.g., a reactor trip) or, if it is complicated by equipment malfunction or operator error, will indicator sed by NRC risk analysts outside of the process described herein.
Obiectives
- 4. To characterize the risk significance of an inspection finding consistent with the regulatory response thresholds used for performance indicators (Pis) in the NRC licensee performance assessment process and for entry into the enforcement process.
5.Tobnce of inspection findings. vide a risk-informed framework for discussing and commu) signi
]
Definina Characteristic Themostimportantcharacteristicof thisprocessisthatitelevatespotentiallyrisk-significantissues early in the process and screens out those findings that have minimal or no risk significance.
Further, field inspectors and their managers should be able to efficiently use the basic accident scenario concepts in this process to categorize individual inspection findings by potential risk significance. The process presumes the user has a basic understanding of risk analysis methods.
Introduction The proposed overail licensee assessment process (as defined outside of this document) evaluates licensee performance using a combination of Pi and inspections. Thresholds have been established for the Pls, which, if exceeded, may prompt additional actions to focus licensee and NRC attention on areas in which there is a potential decline in licensee performance. The inspection finding risk characterization process described in this appendix and illustrated in Figure 1 evaluates the significance of individual inspection findings so that the overall licensee performance assessment process can compare and evaluate them on a significance scale similar to the PI information. Licensee-identified issues, when reviewed by NRC inspectors, are also candidates for this process.
Issue Date: 05/07/99 Al-1 DRAFT 06XX
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Inspectionfindin 11tedtoreIciorsafetycom:rston:s initiltingsv nts,mitigitingsyst:ms,and occupational exposure, public exposure, y thin tha (r:mtining cr: s (em barrier integrity)g3r:will be essessed differentl and physical secunty).
For the reactor saf comerstones, excluding the EP area, each finding is evaluated using a risk-informed framewo that relates the finding to specific structures, systems, or components (SSCs), identifies the core damage scenarios to which the failure of the SSCs contribute, estimates how likely the initiating event for such scenarios might be, and finally determines what capability would remain to prevent core damage if the initiating events for the identified scenarios actually occurred.
BRatt The approach described in this, Appendix was developed using input derived from other agency documents, including the following:
- Regulatory Guide 1.174,"An Approach for Using Probabilistic Risk Assessment (PRA) in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis;"
- Table 1 was based on generic values obtained from NUREG/CR-5499, " Rates of Initiating Events at U.S. Nuclear Power Plants: 1987 - 1995;"
- The accident sequence precursor (ASP) screening rules as outiined in NUREG/CR-4674,
" Precursors to Potential Severe Core Damage Accidents."
In addition, Table 2 is based on generic numbers that are generally consistent with values obtained from PRA models.
Sensitivtv Test of Inspection Findina Risk 1haracterization Screenino Process The staf performed a simple test o1 the sene tivity of the screening process. The test was designed to ensure that items with proven risk importance would not be screened out by the process. The staff reviewed the 1996 accident sequence precursors (ASP) to potential severe core damage events. In 1996, the NRC identified in NUREG/CR-4674, Vol. 25,14 precursors with a conditional core damage probability CCDP greater than 1E-6 affecting 13 units. There were seven precursor events involving in(itiating) events at power, six precursor events involving urmailabilities at power, and one precursor event involving an initiating event while the plant was shut down. With the exception of the shutdown event, which the IFRCP does not currently model, all of the risk significant ASP events successfully passed the screening test and would have required further evaluation using Phase 2 of the model. Because of the simplicity of the model, the process has the potential to overestimate the risk significance of some events, possibly requiring a more refined evaluation before a final characterization can be made.
Process Discussion The inspection finding assessment process is a graduated approach that uses a three-phase process to differentiate inspection findings on the basis of their actual or potential risk significance.
Findings that pass through a screening phase will proceed to be evaluated by the next phase.
Phase 1 -
Definition and initial Screening of Findings: Precise characterization of the finding and an initial screening-out of low-significance findings Phase 2 -
Risk Significance Approximation and Basis: Initial approximation of the risk significance of the finding and development of the basis for this determination for those findings that pass through the Phase 1 screening Phase 3 -
Risk Significance Finalization and Justification: As-needed refinement of the risk significance of Phase 2 findings by an NRC risk analyst Phases 1 and 2 are intended to be accomplished primarily by field inspectors and their first-line managers. Until a user becomes practiced in its use, it is expected that an NRC risk analyst may be needed to assist with some of the assumptions used for the Phase 2 assessment. However, after inspection personnel Decome more familiar with the process, involvement of a risk analyst is expected to become more limited. The Phase 3 review is not mandatory and is only intended to confirm or modify the results of significant (hods will utilize c)urrent PRA techniques and Phase 2 assessment. Phase 3 analysis met the expertise of knowledgeable risk analysts.
NOTE: it is expected that a sample screening and Phase 2 work-sheet, intended for inspector use to aid in their significance determination, will be developed. This work-sheet, along with plant or Issue Date: 05/07/99 Al-3 DRAFT 06XX
design specific information to sJd tha inspectors in using Tcbis 2 of this eppendix will be tran: mitt::d to the pilot plants aftir dev:lopm:nt. Th3 use of this work-cheet will not be m:ndatory, sinc 3 some simple screening may be done mentally. However, for any issue where a Phase 2 analysis is required, the information necessary to reconstruct the Phase 2 analysis needs to be documented in the inspection report since the significance detennination and all assumptions used has to be communicated to the licensee and others who read the inspection report.
Step 1 - Definition and initial Screening of Findings Step 1.1 - Definition of the inspection Finding and Assumed Impact it is crucial that inspection findings be well defined in order to consistently execute the logic required by this process. The process can be entered with inspection findings that involve one or more degraded conditions concurrently influencing safety - or non-safety-related equipment and/or initiating event frequency. The definition of the finding should be based on the known existing facts and should NOT include hypothetical failures such as the one single failure assumed for licensing basis design requirements. - The statement of the finding should clearly identify the equipment potentially or actually impacted, as this will be used in the nsk characterization process, in some cases, the impact of the finding can be stated unambiguously in terms of the status of a piece of equipment, for example, whether it is operable or not, or whether it is available to perform Its function or not. In other cases, the finding may s i
equipment becomes unavailable. In still other cases,pecify conditions under which a piece o exampi,e, the impact is not determined, and assumptions will have to be made for the purposes of assessing the risk significance. Any expliclity stated assumptions regarding the effect of the finding on the safety functions should initially be conservative (i.e., force a potentially higher risk because the final result will always be viewed from the context of those assumptions.
significance)information or analysis from the licensee or other sources is expected, in ma Subsequent to reduce the significance of the finding, with an appropriate explicit and defensible rationale.
Findings must also be well defined because the assumptions can be modified to examine their influence on the results. However, the general rule is that the definition of the finding must address the follow, unction impact and any assumptions regarding other plant conditions. Examples include itssafetyf ing:
The following situations represent two different findings: a motor-operated valve (MCV)ith 1.
in a pressurind-water reactor (PWR) auxiliary feedwater (AFW) system is found w hardened gearbox grease (i.e., is degraded); and an MOV in the AFW system is found with a broken wire that renders it non-functional. For the purposes of assessing the risk significance, the impact of both could be characterized conservatively as "MOV does not perform its safety function of opening to provide flow to the steam generators." In the first case, it is necessary to assume that the hardened grease makes the valve unavailable, while in the second it is not.
2.
A finding involving a deficiency in the design of the plant could be stated as follows:
" Equipment / System / Component X would not perform its safety function of.... under conditions.... For example, a remote shutdown panel that mightbe rendered inhabitable during a cable spreading room fire that causes a loss of offsite power due to inadequate i
heating, ventilation, and air conditionin (HVAC) dispersion of the resulting smoke, would i
be characterized conservatively as.
nt cooldown not ible from centrol room or remote shutdown panel during a loss offsite power (L P) caused by cable spreading room fire due to inhabitability from resulting smoke and loss of power to remote shutdown panel HVAC."
Stop 1.2 -i.74tial Scrooning of the inspection Finding For the sake of efficiency, the initial screening is intended to screen out those findings that have minimal or no impact on risk early in this process. The screening guidelines are linked to the comerstones as follows: If there is negligible impet on meeting the reactor safety comerstone objectives, the finding can be identified as having minimal or no impact on risk and should be corrected under the licensee's corrective action process.
The decision logic is described as follows:
If the finding and its' associated assumptions, as defined in Step 1.1, could simultaneously adversely affect two or more reactor safety cornerstones, then Phase 1 is complete and the user should proceed directly to the Phase 2 analysis. Altematively, the finding can be screened out
- 06XX DRAFT Al-4 Issue Date: 05/07/99
immedilt:ly (chirret2rized es having littl3 or no ri:k pot:nti:1 imp;ct and exit this proc:ss) if it can be shown to NOT be r: lit:d to cny advarse eff:ct on any recetor saf:ty comerstons. Finally, if the finding and its associated assumptions affect only ONE reactor safety comerstone, it may still be screened out as follows:
If only the mit'gation systems comerstone is affected and the finding and the associated i
assumptions do NOT represent a loss of safety function of a system, OR the finding and associated assumptions represent a loss of safet for operation (LCO for Technical Specification (y function of system for LESS THAN the allowed outage time AOT) prescribed by the limiting condition equipment, OR represents a design or qualification findi but the equipment or the system is still operable (e.g., meets NRC I
Generic Letter 91-criteria to remain operable), OR is not categorized as a risk-significant SSC under ue maintenance rule (10 CFR 50.65) then the finding would be considered green and screened out.
If only the initiating event comerstone is affected and the finding and associated assumptions have no other impact than increasing the likelihood of an uncomplicated reactor trip, the finding would be considered green and screened out.
If only the fuel barrier is affected, the issue will be screened out since a Pl exists for this barrier.
If any reactor coolant system (RCS) barrier function to mitigate an accident sequence is affected, the issue will be assessed in Phase 2.
If the containment barrier is affected, the concem is referred to a risk analyst until more guidance can be provided. However,if the concern is associated with containment cooling function needed to preserve the NPSH capability of the ECCS equipment during the recirculation phase, its impact should also be evaluated as part of mitigation system comerstone above.
Any inspection finding that is NOT screened out (i.e., characterized as green) by the above-mentioned decision logic should be assessed using the Phase 2 process described herein.
l Phase 2 - Risk Significance Approximation and Basis Stop 2.1 - Define the Applicable Scenarios Once an inspection finding passes through the Phase 1 screening, it is evaluated in a more detailed manner using the Phase 2 process described herein. The first step in Phase 2 is to ask the question "Under what core damage accident scenarios would the finding, as defined in Step 1.1, increase risk?"
Determinin Therefore, g which scenarios make an inspection finding risk important may not always be intuitiv documents such as plant-specific PRA studies, safety analysis reports, TS bases, and emergency operating procedures should be reviewed as needed to ensure that the most likely events and circumstances are considered. Specifically, the inspector must determine which core i
damage scenarios are adversely impacted by each finding.
. Identifying the scenarios begins with identifying the equipment and the assumed or actual impact of the finding, and takes into consideration the role the equipment plays in either the continued operation of the plant or the response to an initiating event. This step leads to an identification of the role of the findin in either contributing to an initiating event or affecting a mitigating system, or both. For the miti ating systems, the impact may be one of two kinds: the finding results in the equipment function' being compromised or the findin under which the function would become compromised.g relates to the identification of a cond In the first of these two cases, the function can be assumed to be lost, and the scenario of interest is the initiating,de the same function event for which the equipment is required and the remaining equipment that by design can provi as that which has been lost. For the second case, the scenario definition must also include the condition under which the function would become compromised. For example, if the finding is that if two operator actions are reversed while performing the rwitchover to recirculation in a PWR, the safetyinjection (SI)f coolant accident (LOCA) initiating event, the failure of the charging syste pumps could be irreparably damaged due to cavitation, the scenario definition includes the loss o it is a viable altemative means of providing sump recirculation, and also the human error represents the condition under which the pumps would fail). If)the finding were that the Si p(w umps Issue Date: 05/07/99 Al-5 DRAFT 06XX
a point only), tha scenino d 2finition would involva(only ths LOCA and th3 charging During this phase of the process, inspectors may determine that several different scenarios are affected by a particular inspection finding. This determination can occur in one of two ways:
First, the finding may be related to an increase in the likelihood of an initiating event, which may require consideration of several scenarios resulting from this initiating event.
Second, a finding ma events. Forexample,y be related to a system required to respond to several initiat response to both a loss of offsite power and a LOCA. Each of these two initiating events must oe considered separately so that the next step of the Phase 2 evaluation process can determine which scenano is potentially most significant.
The scenario resulting in the highest significance will be used to establish the initial relative risk-significance of the finding. If a Phase 2 assessment of multiple applicable scenarios results in all
" green" nificance, the user should seek assistance of a risk analyst, since the Phase 2 process cannot
" sum" the significance of multiple low-significance scenarios. Additionally, a particularin finding may affect multiple comerstones by both increasing the probability of an initiating event and degrading the capability or reliability of a mitigating system. Again, each applicable scenario must be considered to determine which is the most significant.
In identifying possible core damage accident scenarios, consideration must also be given to the roleof systems as well as the primary system. For example, if a particular initiating event can be m edby more than one system providing the same safety functior.. but all such systems are depe nt on a single train of a support system the limiting scenario may involve the failure of the sing (e.g., service water or emergency ac individual primary system trains.
Step 2.2-Estimation of the Likelihood of Scenario initiating Events and Conditions in St 2.1, sets of core damage accident scenarios were determined that could be made more likel the identified inspection finding (degraded condition). This step should result in the tion of one or more initiating events, each followed by various sequences of equipment failures or operator errors. To determine the m ost limiting scenario, perform the following analysis for gash set of scenarios with a common initiating event.
If the finding does not relate to an increased likelihood of an initiating event, the initiating events for which the affected SSC(s) are required are allocated to a frequency range in accordance with guidance provided in the left-hand column of Table 1 herein. Table 1 is entered from the left column, using the initiating event frequency, and from the bottom, using the estimated time that the degraded condition existed, to arrive at a likelihood rating (A - H) for the combination of the initiating event and the duration of the degraded condition.
If the finding relates to an increased likelihood of a specific initiating event, the likelihood of that initiating event is increased according to the significance of the degradation. For example, if the 1
inspection finding is that loose parts are found inside a steam generator, then the frequency of a I
for that steam generator tube rupture (SGTR)dingly. plant may increase to the next higher frequencj category, and Table 1 is entered accor When the scenario includes the identification of a condition under which a function, a system, or a train becomes unavailable, then this fact has to be factored into the assessment, it is not appropriate to assume that the affected function, system, or train is unavailable. At this point, it is necessary that a risk analyst assess the probability of the condition, and adjust the likelihood of the initiating event (or events) by the appropriate amount. For example:
- A finding that if a control valve in the instrument air system fails it could lead to overpressure of a low-ressure part of the system, thereby leading to the failure of the equipment controlled b the air system. The probability of interest is that of the failure of the valve during the mi ion time, which depends on the impact of the failure. For example, if the valve failure would lead to a reactor trip in addition to failing some mitigating equipment, the mission time is 1 year, and the initiating event frequency would be the j
probability of failure of the valve in one year. If the impact is simply on the mitigating systems for a LOCA, the mission time is that time required to place the plant in a safe, i
06XX DRAFT Al-6 Issue Date: 05/07/99
r stable stat 2.
In this case, the LOCA frequency would be cdjusted by th3 probability that the valve f;ilure would occur during the mission time.
Finally, remember that the definition of the finding and the selection of core damage accident scenarios should be strictly based on the known existing facts and should NOT include hypothetical failures, such as the one single failure assumed for licensing basis design requirements.
Issue Date: 05/07/99 Al-7 DRAFT 06XX
Table 1 - Estimated Likelihood Rating for Initiating Event Occurrence During Degrcded Period (taken from NUREG/CR-5499)
Row Approx. Freq.
Example Event Type Estimated Likelihood Rating
>1 per 1 - 10 yr Reactor Trip A
B C
I Loss of Power Conv. Sys.
(loss of condensor, closure of MSIVs, loss of feedwater) 2 1 per 10 - 10 yr Loss of Offsite Power B
C D
MSL(B (outside entmt) Stuck open SRV only) 11 2
1 per 10 - 10 yr SGTR C
D E
Small LOCA (PWR) lli (RCP seal failures and stuck open SVs only)
MFLB MSLB (inside PWR cntmt) 1 per 10 - 10' yr Small LOCA (pipe breaks D
E F
ATWS-PWR (elect only) )
IV 1 per 10'- 10 yr Med LOCA E
F G
5 V
Large LOCA ATWS-BWR (BWR) 5 Large LOCA (PWR F
G H
<1 per 10 yr ATWS-PWR (mech)only)
VI ISLOCA I Vessel Rupture
> 30 days 30-3 days
<3 days Exposure Time for Degraded Condition Table 1 - Estimated Likelihood for initiating Event Occurrence During Degraded Period Use of Table 1 should result in one or more initiating events of interest with an associated likelihood rating ("A" through "H") for each.
Step 2.3 - Estimation of remaining mitigation capability The scenarios of interest have now been identified, and Table 1 has been used to estimate associated initiating event frequencies and to combine them with degraded condition exposure time to arrive at an estimate of the likelihood of the initiating events. Following an initiating event, core damage will result from a series of system, component, or operator failures. In this step, the user will approximate the robability of faifing to mitigate the coro damage scenarios associated with the condition identified the finding. Findings defined in Phase 1 will generally identify the potential for degrading a part lar function. Therefore, the probability of, preventing the scenarios that the function. graded function will depend on the number of remaining success paths for providing include this de To count success paths in a probabilistically consistent manner, systems are considered to be either single train or redundant. A redundant system is a system that has more than one identical train, where the loss of one train does not lead to a loss of function. However, all trains of a redundant system are subject to a possible common-cause failure. Success paths may be 06XX DRAFT Al-8 Issue Date: 05/07/99
provided by each trdn of div rs3 cingl:-train syst:ms (.g., high-pressura injection in a boiling wat;r r::ctor (BWR) for a loss of fredwit:r transh nt mi provided bytha h 1-pressuracoolant injection HPCI and reactor core isolation coolant RCl systems, both si train systems), or by divers (e redu)ndant systems (e.g., low-pressure i(njection may be provi the low-pressure core spray (LPCS)dundant systems. In addition, in some cases there may be ime to and the LPCI systems in a BWR-4, both multi train system, or by mixtures of single-train and re function or train that has been lost, which can be credited as a success path under certain conditions.
{
'In counting the number of remaining available success paths for a scenario affected by the 2, gradation assumed by the finding, the user must select the most appropriate column of Table de Risk Significance Estimation Matrix," for each affected scenario. Each column in Table 2 represents about one order of magnitude difference from adjacent columns in the failure probability of remaining success paths, and the descriptions in the column headings are intended as examples of mitigation methods that can typically be assumed. Refer to Figure 2 for basic guidance on how to determine the number of trains and redundant systems. In addition, the following rules and guidelines apply:
Only equipment that the licensee has scoped into the maintenance rule (10 CFR 50.65) may be credited for remaining mitigation capability. This provides a minimum level of I
assurance that credited equipment meets pre established reliability goals or performance criteria.
j The pottotial for common-cause failure of the romaining success paths is accounted for in the column definitions of Table 2. Therefore, any actual evidence of a common-cause failure must be included in the definition of the inspection finding.
Credit for recovery may be taken if there is a possibility of restoration of the SSC or a function that has been assumed to be lost due to the condition identified by the finding.
Recovery actions should be credited only if there is sufficient time available, environmental conditions allow access, they are covered by, operator training and written procedures, and necessary equipment is available or appropnately staged and ready. For recovery actions that are relatively complex, and/or require actions outside the control room, it is particularly j
important that the actions required are feasible within the time available to prevent core damage. If there are no remaining success paths other than restoring the failed equipment, and the above conditions are met, then Column 6 of Table 2 will credit this recovery. For example, consider an inspection finding involving a potentially recoverable system failure, such as a, failed automatic start feature. If status indication exists and simple operator action would be able to start the equipment within sufficient time to provide the system function, then more credit can be given to recovery, which may be more appropriately given by using Column 5. If other equipment is also available as remaining success paths, then operator actions may be used to supplement that equipment.
Caution has to be exercised when taking(credit for systems that are dependent on ma actuation (such as standby liquid control SLC) in BWRs), if the time to initiate the system is short and performed under stressful conditions, Column 5 should be used for a redundant system rather than Column 4. When there is ample time, as in the initiation of suppression pool cooling in BWRs, the human error probability is low enough that the nominal system column can be used.
When all scenarios have been assigned and the associated likelihood and remaining mitigation capability estimated, the Table 2 matrix described in the next section can be used to estimate the potential significance of the degraded condition, within the context of all assumptions made to this point.
)
Step 2.4 - Estimating the Risk Significance of inspection Findings The last step of the Phase 2 assessment process is to estimate the relative risk sig)nifican finding. The risk is estimated by employing an evaluation matrix Table 2 herein, which utilizes the information gained from Steps 2.1 through 2.3. This matrix co(mbines the scenario like derived in Step 2.2 with the remaining mitigation capability determined in Step 2.3 and establishes an estimated risk significance for the particular finding. One of only four possible results can be obtained: Green, White, Yellow, or Red. These results are comparable to those used for Pls. The user must complete this assessment process forpagh scenario affected by the inspection finding i
before determining the scenario of highest significance.
j Issue Date: 05/07/99 Al-9 DRAFT 06XX
As e mental " benchmark," th3 user of this process should recogniz3 that a " Green" outcoms will involve any condition th;t has three or mora div rse irgins of remrining mitigation capability no j
matter how frequently it oxurs, and that a " Red" outcome will involve any condition that.has zero 1
or only one train of remaining mitigation capability if the initiating events that require such capability I
occur more often than once every 1000 reactor-years (e.g., a small LOCA, a LOOP, or a reactor trip).
i 06XX DRAFT Al-10 Issue Date: 05/07/99
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I Step 2.5 - Documenting the Results
(
l The results of the Phase 2 risk estimation will be communicated to the licensee through the inspection i
report process. It is expected that risk-significant or controversial findings will require obtaining l
licensee risk perspectives and will most likely prompt a Phase 3 review. If the inspectors, and appropriate regional and Headquarters staff (when necessary), a, gree with the results of the Phase 2 assessment, the final results will be documented in an inspection report and no further review is needed. The extent of documentation should include allinformation needed to reconstruct the Phase l
2 analysis. Although licensee perspectives will be considered, the NRC staff will retain the final responsibility for determining the risk significance of a finding and will provide its justification in an inspection report or other appropriate document. When licensee assumptions or perspectives differ from those of the staff, the staff should explicitly justify the basis for its determination.
l Phase 3 - Risk Significance Finalization and Justification if determined necessary, this phase is intended to refine or modify the earlier screening results from 1
Phases 1 and 2. Phase 3 analysis will utilize current PRA techniques and rely on the expertise of l
knowledgeable risk analysts. The Phase 3 assessment is not described herein.
l l
l l
l 06XX DRAFT Al-12 Issue Date: 05/07/99
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Appendix 2 Significance Determination of inspection Findings in the Emergency Preparedness, Radiation Safety, and Safeguards Area This appendix and its attachments represent the concepts for evaluating inspection findings in the emergency preparedness, radiation safety, and safeguards areas. Thresholds were selected on a significance scale similar to those established for the plant performance indicators that industry plans to submit.
I i
Issue Date: 05/07/99 A2-1 DRAFT 06XX EMERGENCY PREPAREDNESS SIGNIFICANCE DETERMINATION PROCESS The objective of this comerstone is to ensure that the licensee is capable of implementing adequate protective measures to protect public health and safety in the event of a radiological emergency.
Licensee performance in the comerstone is assessed by considering both the relationship of pedormance indicators. (PI's) dispositions individual inspection findings in a manner that allo with regard to thresholds and inspection findings. The Significance DeterminationProcess SDP) to be combined with PI(results.
The SDP consists of flow chart logic to disposition inspection findings into one of the following categories: " licensee response band,"" increased reg"ulatory response band,"" required regulatory response band," or " unacceptable performance band.
During the development of Emerg,ency Preparedness (EP) PI's, the most risk significant areas were identified as distinct from other important program elements. These development efforts were performed by a group of EP subject matter experts with input from members of the public. The SDP methodology recognizes failures in the identified risk significant areas as more significant than findings in other program areas.
Emergency Pr redness regulations codify a set of emergency planning standards in 10 CFR 50.47,(b) and ndix E to Part 50. The more risk significant areas of EP align with a subset of the plannmg sta s and requirements. The SDP logi standards and other regulatory requirements as criten,c uses failure to meet or implement plaI the more risk significant planning standards results in greater significance (e.g., a white finding as opposed to a green finding.)
The logic of the SDP intentionally parallels NUREG-1600, " General Statement of Policy and Procedures for NRC Enforcement Actions." The GREEN, WHITE, YELLOW and RED results genera!ly ah'gn with current Severity Levels IV, Ill, ll and I respectively. However, there are some differences that were generated as a result of subject matter expert efforts to identify the most risk significant areas of EP. The SDP does not sum unrelated findings to escalate the resultant response band disposition. However, a program failure may be indicated by contemporaneous failure to meet multiple planning standards. The SDP logic recognizes this unlikely, but significant, deterioration of an EP program and responds with findings of increased significance, including the potential for a set of concurrent findings being assessed as " unacceptable performance."
A finding that is assessed as a GREEN indication does not mean that the performance associated with the finding is good or even acceptable. It may represent non-conformance or a violation.
However, the safety, significance of the finding is not great enough to warrant further NRC intervention. It is considered to be within the " licensee response band. Ucensees are still required to retum to compliance with the regulation and their commitments. However, the licensees are given the latitude to self correct these findings.
l 06XX DRAFT A2-2 Issue Date: 05/07/99
Guidance 1.
Exercise performance is measured by the Drill / Exercise Performance and ERO Drill Participation performance indicators. Exercise weakness and deficiencies are expected to be identified and resolved by the licensee in accordance with 10 CFR 50.47(b)14. The inspection program is designed to verify that this expectation is met. However, poor performance itself is not a finding.
2.
A Finding is an observation of an emerge,ncy preparedness program element that has been placed in context and assessed for significance.
3.
Failure to implement a planning standard means that it was not implemented during an emergency event, but that the program itself continues to meet the planning standard e. g.,
a personal error during an event.
4.
Failure to meet a planning standard means that the program is not in compliance with the planning standards of 50.47(b) and the requirements of Appendix E as committed to in the site Emergency Plan. The measure of program compliance are the criteria of NUREG-0654, as articulated in the approved Emergency Plan.
5.
A regulatory requirement is any requirement of 10 CFR or the Emergency Plan.
6.
A violation of requirements may also involve a failure of the PIDR. This should be analyzed through the significance determination process for both the violation and the PIDR failure. The more significant determination should be the overall determination.
7.
Failures of the PIDR program that result in green, white or yellow findings should also be provided to the inspection team responsible for the conduct of Inspection Procedure No.
71152 Identification and Resolution of Problems.
Issue Date: 05/07/99 A2-3 DRAFT 06XX
NRO Sg' nincarce Determination Process for Emergency Pamparedness inspection Fhdngs Sheet 1 drat 3 April 19, t999 1
Finding identified i r hd YES
,gg
- ToViolaiian 9
Shed 2 NO 1 r N[
> TOPIDR Problem Shed 4 NO i f inepecilon cheervenien l
REFS =The RiskSignincant Planning Standards 80.47 (b) 5,4,9 8.10 and Apperdk E section IV B, C, D(t) Ik D(3)
PS = The Planning Standards of 50.47 (b) and the requirements d Apperdk E PDR = Poblem identification and Resoknen System Trnely Resolution d failure to meet = RSPS, so Day; PS,120 Day; other Regulatory Requirement,240 Day j
Trnely Resolution d failure to implement = RSPS, immediate (t 4 Day); PS, to Day; cther Regulatory Requirement, t20 Day i
06XX DRAFT A2-4 Issue Date: 05/07/99 o
NRC Significance Det:rmination Process for Emergency Preparedness inspection Findings Sheet 2 draft 3 April 19,1999 l
i
[ Violation identifiedD N
)l l
1 P Failure to NO Irnplernent or 7
GREEN meet PS?
YES 1 F Failure to YES Igm
- t Pk?
O To PS Implementation Problem Eq Sheet 3 NO 1 P FaHure to NO 5 or more NO
(
ure)
I
M U
MM YES YES 1 f 1 F 3 or rnore NO i
failures m meet YELLOW RSPS?
YES 1 r RED Issue Date: 05/07/99 A2-5 DRAFT 06XX
NRC Significance Determination Process for Emergency Preparedness inspechon Findings - Sheet 3 draft 3 April 19,1999 PS Implementation Problem (Actual Event) 1 r NOUE YES
?
7 GREEN NO 1 r ALEpg YES FaAure to YES >
WHITE 7
knplement RSPS?
NO 3
GREEN 1 r SAE YES Failure to YES Ignt 7
YELLOW 7
NO j
7 WHITE NO l f GEN YES Failure m YES EMERG U
nt 7
RED y
NO 3
YELLOW 06XX DRAFT A2-6 Issue Date: 05/07/99
NRC Significance Determination Process for Ernergency Preparedness inspection Findings Sheet 4 draft 3 April 19,1999 PIDR Problem i
1 P PIDR FMLURE YES Dr#VExerase To PlDR Drill / Exercise Ev Sheet 5 l
Inspection Obsenrenon g
4 k 1 V Failure Failure e Fe4m 2
- N*
NO IDPS NO ID other Of f Probiern?
regulatory a
(rnest or regarernent
,#mg*
implemen0
?
YES YES YES q 7 1 P GREEN l
l 1 P yg PS7 WHITE 1
l t
i NO 1 P l
Other yg$
GREEN req nts l
l NO 1
P inspection Observaton i
Issue Date: 05/07/99 A2-7 DRAFT 06XX
7.-
NRC Significance Determination Process for Ernergency Preparedness inspection Findings - Sheet 5 draft 3 April 19,1999 hlDR Drill / Exercise Evaluation Problem 1 r Fgto f$[sp$
MI YES YES YES
'T T
?
?
?
NO I f NO 1 r Repeat YES Fai re 3
WHRE
?
NO NO 1 P GREEN 1 P Failure to ID regulat NO he om
@Y YES 1 r GREEN 06XX DRAFT A2-8 Issue Date: 05/07/99
r
)
OCCUPATIONAL RADIATION SAFETY SIGNIFICANCE DETERMINATION PROCESS The objective of this comer stone is to ensure worker health and safety from exposure to radiation from licensed or un-hcensed radioactive materials during routine operations of civilian nuclear reactors. The health and safety of workers is assured by maintaining their doses within the limits in 10CFR20 and ALARA.
Licensee performance in the cornerstone is assessed by considering the Pi indication in combination with inspection findings. A baseline inspection is maintained to verify the accuracy and completeness of the Pi data (i.e., work control in radiologically si sufficient to measure performance (gnificant areas), supplement the Pl data in areas where the Pl al i.e., problem identification and resolution and complement the Pls with inspection findings of performance for areas not covered by the Pi (i.e., ALARA), planning and cont monitoring instrumentation, and personnel dosimetry).
The Significance Determination Process (SDP) is the mechanism in which the significance of individual events I
(follow-up of an rational occurrence, substantiated allegation, or other inspection finding) can be normalized
]
cnd combined ' the PI results to arrive at an overall comerstone performance assessment. Logic flow charts tre provided to outline the process. A finding that gets through the process (flow chart)is individ withoutinpping a decision
" gate" ends up as a GREEN indication. This does not mean that the performance on th or even acceptable. It still may be a non-conformance or a violation, it does mean that the safety significance i
of the event is not large enough to warrant further NRC intervention. Licensees are still required to come into compliance with the regulation and their commitments. However, the licensees are given the latitude to celf correct these non-conformancec.
The decision gates in the SDP intentionally parallel the Enforcement Policy Supplement examples to facilitate use Althoug(h not a fast rule, the GREEN, WHITE,pection findings tha of the SDP in evaluating the significance of ins l
YELLOW and RED results generally align with current Severity l
Levels SL) IV, Ill, ll and I respectively. In addition, there is considerable overlap with the Performance Indicator
. definitions to allow the SDP to be used to determine those single events that, although they meet the Pi definition l
(tnd will be counted as a PI), they require separate reporting by Part 20 and may be a significant enough risk to worker health and safety to deserve their own " colored" assessment input.
l ALARA Findinas l
Section 1101. b) of 10 CFR Part 20 states that licensees shall use, to the extent practical, procedures and l
engineering co(ntrols based upon sound radiation protection principles to achieve occupational d l
l low as is reasonably achievable (ALARA).
An ALARA finding is a finding whereby the licensee has failed to properly implement procedures and engineering controls based on sound radiation protection principles to ensure that doses associated with plant operations and miintenance are maintained ALARA.
l protection program that includes provisions for keeping occupational radiation doses ALARA.ple Section 1101 of 10 CFR Part 20 reguires that each licensee develop, document, aled im l
l As containedin the Statements of Consideration in the May 21,1991 Federal Register conceming the revision to 10 CFR Part 20, the l
Commission continues to emphasize the importance of the ALARA concept to an adeguate radiation protection program. A licensee's compliance with this requirement will be judged on whether the hcensee has incorporated m:asures to track and, if necessary, to reduce exposures and not whether exposures and doses represent an cbsolute minimum or whether the hcensee has used all possible methods to reduce exposures.
i The metric chosen for the ALARA portion of the SDP for evaluating the significance of an ALARA finding is a plant's three-year averaoe collective dose consideration of doses to individuals and individual dose limits are tr:ated in theE.xposure Control portions of(the SDP). Plants with effective ALARA programs tend overall collective doses than those which have poor or inadequate ALARA programs. On average the industries current ALARA performance is considered very good. Total collective dose appears to be reaching an equilibrium minimum value in the last few years. Therefore, current the median value of the three-year average (MTYA) collective dose and the third quartile values are established as a decision gate standards. Due to the different challenges for BWRs and PWRs, different MTYA and third quartile values are established for these reactor types.
Issue Date: 05/07/99 A2-9 DRAFT 06XX
Another matric which is used to evalu2 tid ALARA findings is ths accuracy of a lic nsse's dose goals establish d for work packag:s. A job doss which exc:eds the dose goal by 50% or more is indicative of poor pre-job planning cnd one proceeds to the next gate. If the actual job dose falls within the pre-job dose estimate or exceeds it by 1:ss than 50%, then the finding is GREEN.
The next two gates evaluate the si nificance of the ALARA finding once the magnitude of the actual job dose has exceeded the job dose estimate b 50% or more. Once the magnitude of an actual job dose reaches 20% of the MTYA for the appropriate reactor
, then the significance of the ALARA finding goes from GREEN to WHITE.
The occurrence of three or more doses which are greater than 4% but less than 20% of MTYA (in a one calender year period) will also give a WHITE finding.
The final gates in the ALARA flowchart use a plant's source term as a metric and evaluate whether the finding is a source control finding or not. If a plant is a known high source term plant, then the licensee should be sensitive to this issue. The licensee should place increased emphasis on ensunn,g that radiation fields in the work area are minimized. If a plant is a high source term plant and finding is determined to be a source term finding (i.e., the high job dose resulted from inadequate pre-finding will be classified as being YELLOW. job planning to reduce the radiation fields in the work are If the finding is not determined to be a source term finding, then the finding will be classified as being WHITE.
Similarly, if the plant is not a high source term plant, then more emphasis should be placed on work controls such cs proper job planning, use of mockups, use of skilled workers, integration and coordination of jobs to minimized unnecessary setup and tear down of scaffolding and temporary shielding, etc. If a finding is identified (for a plant which is not a high source term plant) in which the high, job dose was a result of lack of attention to work controls (i.e., it is not a source control problem), then the finding will be classified a being YELLOW. If the finding is datermined to be a source term finding, then the finding will be classified as being WHITE.
Exposure Control Findinag Unintended sure is a failure of one or more radiation safety barriers that results in a significant unintended or unplanned
. Radiation safety barriers include adequate radiation monitoring, physical controls, hazards analysis and surveys, instructions to workers (including RWPs, postings, Part 19 training, etc) and proper, job controls. Since most of these are either required by 10 CFR or covered in technical specifications required procedures, the failure to implement one or more of these is generally a viciation. To be significant, the consequence (unintended dose) needs to equal or exceed a set percentage of the dose limits in 10 CFR 20 ( 2%
of the stochastic,10% of the non-stochastic,20% of the limits to minors and fetus, and 100% of the shallow dose limit from a Discrete Radioactive Particle).
Exposure in excess of 10 CFR 20 limits is a YELLOW indication. This is an area where SDP deviates from the Enforcement Policy (SL lli" equate" to WHITE findings). Based on the ICRP 66 total probability coefficient of 4E-4
/ rem, a 5 rem dose equates to a 2E-3 risk of death from cancer. Since preventing this risk is the objective of the Ridiation Safety comerstone, we have elevated its significants from that in the Enforcement Policy. The threshold for a RED finding is consistent with the SL I significance in the Enforcement Policy.
Breakdowns in the Radiation Protection Program, or unintended exposures, that do not exceed a dose limit can still be considered significant if they constitute a " Substantial Potential for Overexposure." A substantial potential, consistent with the current Enforcement Manual NUREG/BR-0195, subsection 8.4.1), is an occurrence in which a minor alteration of the circumstances would hav(e resulted in a violation of Part 20 lim ths altered circumstances did not occur. In the SDP the finding can also be a YELLOW or RED depending on l
th3 dose rates (risk of a serious outcome) associated with the failure. In a Very High Radiation Area of 500 l
rads /hr, it can take as little as 3 minutes for a worker to receive 25 rem. Note however that the Enforcement Process En actual (and possible civi_l penalty) will not engage unless the event had an " actual consequence" (in this c overexposure). The Assessment Process rather than the Enforcement Process will determine further licensee and NRC action for events that do not result in " actual consequences."
The last decision gate in the SDP is intended to sort out significant issues and findings related to plant equipment and facilities. The Assessment Program is a risk informed process, and radiation dose is the measure of health risk associated with licensee activities. Therefore, this gate focuses on those issues that could or does compromise the licensees ability to assess dose. Since this gate culls out WHITE findings (comparable to SL lil),
it is intended that only significant, programmatic, failures of radiation monitoring and personnel dosimetry trip this gite. Examples of findings intended to be addressed by this gate include; 1) the licensee's failure to use a NVLAP certified dosimeter processor,2) a generic and uncorrected failure of the DADS to respond to, or record, radiation dose, and 3 improper calibrations of instruments or monitors that significantly bias their response which j
tra used as a basis)for establishing protective controls. An individual failure to survey or monitor sho 06XX DRAFT A2-10 Issue Date: 05/07/99
considered as a fiilurgof a radiation safety barri r and evaluated.for its potential for unint:nded dose er substantial potential for over;xposura ts discussed above.
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i Issue Date: 05/07/99 A2-11 DRAFT 06XX
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OCCUPATIONAL RADIATION SAFETY Draft ALARA finding 1r Actual N.
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gym l Issue Date: 05/07/99 A2-13 DRAFT 06XX
PUBLIC RADIATION SAFETY SIGNIFICANCE DETERMINATION PROCESS Radioactive Effluent Release Program This branch of the logic diagram focuses on the licensee's radioactive effluent release program.
It assesses the licensees ability to maintain radioactive effluents ALARA. These are the design dose objectives contained in Appendix l to 10 CFR Part 50. Radiation dose to a member of the public is the success enterion.
The regulatory basis for requiring radiological effluent monitoring pr,,ograms is given in General Design Criterion 60," Control of releases of radioactive materials to the environment, of Appendix A," General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50," Licensing of Production and Utilization Facilities." Criterion 60 requires a licensee to provide for a means to control the release of radioactive materials in gaseous and liquid effluents during normal reactor operation, including anticipated operational occurrences. An additional requirement is in Section IV.B.1 of Appendix 1 to 10 CFR Part 50. This section requires a licensee to provide data on the quantities of radioactive matenal released in liquid and gaseous effluents to assure that such releases are within the ALARA design objectives. This data, pursuant to 10 CFR 50.36a, is tyorted to the NRC annually.
SDP determination process: Is there an event or occurrence in the licensee's radiological effluent monitoring program that is contrary to NRC regulations or the licensee's Technical Specifications (TS), Offsite Dose Calculation Manual (ODCM), or procedures? If yes, the question is what is the dose impact (as calculated by the licensee) of the event? If there was no radiological release associated with the event (no dose impact to a mtmber of the public) then there is minimal " risk" and the SDP classifies it as GREEN. The licensee is r:sponsible to resolve the finding. The NRC will periodically inspect the effectiveness of the licensee's corrective tction program.
if the event resulted in an effluent release of radioactive material that, based on the methodology in the licensee's ODCM, exceeded the dose values in Appendix I to 10 CFR Part 50 but is less than 0.1 rem, the SDP classifies ths event as WHITE. In this case, the NRC will maintain some detail of oversight on the licensee's corrective actions. NOTE: The licensee has a Performance Indicator (PI) in this area that uses dose values equal to the quarterly dose values given in the TS or the ODCM. This SDP is not to be used to " double count" the Pl. If a i
situation results in which the dose exceeds Appendix I values because of multiple effluent releases which i
exceeded the Pl threshold it should not automatically be assessed as a degraded comerstone. The SDP is to be used to assess the significance of a finding on an action or event by the licensee which was contrary to NRC regulations, the licensee's TS, ODCM, or procedures.
If the event resulted in effluent release of radioactive material that, based on the methodology in the licensee's ODCM, exceeded the annual public dose limit in 10 CFR Part 20 of 0.1 rem but is less than 0.5 rem, the SDP classifies the event as YELLOW. The NRC would be significant NRC oversight of the licensee's corrective cetions.
If the event resulted in effluent release of radioactive material that, based on the methodology in the licensee's ODCM, exceeded 0.5 rem, the SDP classifies the event as RED. The NRC has lost confidence in the licensee's cbility to control radioactive effluents. Significant NRC interaction with the licensee will result.
Example:
The licensee had an inoperable radiation monitor on the radioactive liquid effluent discharge line. Because the monitor was inoperable, the licensee was required to perform grab sample monitoring of the liquid discharge. The lic nsee failed to perform the sampling, to verify that the liquid effluent was within the activity projected based on prior analysis of the hold-up tank. This is the finding. Looking at the SDP flowchart, the key decision to determine tha significance of the finding is dose. Was the calculated dose from the release above orbelow the values in the decision diamonds? The dose determines the significance color.
Radioactive Environmental Monitoring Program This bra,nch of the logic diagram focuses on the licensee's ability to operate an effective radioactive environmental monitonng program.
- The regulatory basis for requiring radiological environmental monitoring programs is given in General Design Critenon 64, " Monitoring Radioactivity Releases," of Appendix A, " General Design Cnteria for Nuclear Power Plants," to 10 CFR Part 50," Licensing of Production and Utilization Facilities." Criterion 64 requires a licensee to 06XX DRAFT A2-14 Issue Date: 05/07/99
provida for a means for monitorinc tha plant environs for radioactivity that may bs released during normal operations, including anticipated opsrational occurreness, and from postulated accidsnts. An additional requirement is in Section IV.B.S of Appendix I to 10 CFR Part 50. This section requires that the monitoring program identify changes in the use of unrestricted areas (e.g., for agricultural purposes) to permit modifications in the monitoring program for evaluating doses to individuals from principal pathways of exposure.
Rtdiological environmental monitoring accident. During normaloperations,env,is important both for normal operations, as well as in the eve ironmental monitoring verifies the effectiveness of the plant systems used for controlling the release of radioactive effluents. It also is used to check that the levels of radioactive material in the environment do not exceed the projected values used to license the plant. For an accident, the program provides an additional means to estimate the dose to members of the public.
SDP determination process: Is there an event or occurrence in the licensee's radiological environmental monitoring program that is contrary to NRC regulations or the licensee's Technical Specifications (TS), Offsite Dose Calculation Manual (ODCM), or procedures? If yes, the question is; did it degrade the licensee's ability to essess the impact of its radiological effluents on the environment?
To answer the question with a yes means that the licensee's overall program is degraded. It does not mean that a few environmental samples over the course of a year were not taken, or improperly analyzed. A failure in one or two parts of the licensee's program is not sufficient to reach a " White" significance determination. A failure to 6 valuate a required pathway (i.e., no valid data to support an impact conclusion for that pathway) would result in a YES answer to the decision diamond. This is a high threshold to reach. Historically, inspection findings have documented that samples are missed, or a land use census was not for extended periods of time or they were not in the correct location. performed, or the air samplers w data, but not a complete failure to be able to assess the impact on the environment from that pathway. The significance determination of such event would be Green.
Example:
l The inspector observed the collection of air filters from an indicator air sampling station. The inspector discovered that over the previous 12 month period, one of the air sampler was found to be inoperable on 32 separate occasions. This meant that up to 32 weeks of air sample data was missing and/or suspect. Because the monitor was inoperable, the licensee is required to prepare and submit to the Commission, in the annual Radiological Environmental Operating Report, a description of the reasons foi not conducting the program as required and the plans for preventing a recurrence. The licensee failed to prepare and submit the required report. This is the finding. Looking at the SDP flowchart, the key decision to determine the significance of the finding is whether or not the licensee was still able to assess the impact on the environment from radioactive gaseous effluents. In this case the licensee was able to correlate the valid air sample data with known gaseous effluent discharges. Also, th3 licensee had valid air sample data from the sectors on either side of the faulty air sampler. Therefore, the lictnsee had some valid data to use to assess the impact on the environment. Thus, for this case the significance d: termination is Green.
Example:
The in_spector reviewed changes to the radiological environmental monitoring program put in place during the last y:ar. The licensee, based on a review of historical data which showed that no radioactive material of plant origin was detected in any of the fish samples collected in the past 5 five years, eliminated the collection of fish in the river where the discharge canal empties. The inspector identified this as an improper change to the environmental monitoring program because the change reduced the pathway monitoring to below the minimum level acceptable Technical Position on Environmental Monitoring, Revision $, ram is given in the Radiological As to the NRC Guidance for the environmental monitoring,pr complete discussion of the progr,am and changes to the, program over time. The guidance in Regulatory Guide j
4.1 allows the licensee to modify the program after J years of operational monitoring history if it can be i
d monstrated from the levels in environmental media or calculations (using measured effluents and appropriate dispersion and bioaccumulation factors) that the doses and concentrations associated with a particular pathway cra sufficiently small, the number of media sampled in the pathway and the frequency of sampling may be reduced. For this case, the licensee reduced the number of samples and the frequency to zero. Thus, the I
pathway was not monitored. This action compromised the licensee's ability to assess environmentalimpact. The significance determination for this case is White.
Radioactive Material Control Program Issue Date: 05/07/99 A2-15 DRAFT 06XX i
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This branch of tha logic di gram focusts on th3 fictnsas's radioactive material control program. It assess:s the lic:nsto's ability to prev:nt tho inadvortant ralrase of lic:ns:d radioactiva matsrial to an unrestrict:d arca.10 CFR Part 20 contains the requirements for the control and disposal of licensed radioactive material. At a lic nsee's f acility, any equipment or material that came into contact with licensed radioactive material or that had th3 potential to be contaminated with radioactive material of plant origin and are to be removed from the facility must be surveyed for the presence of licensed radioactive material. This is because NRC regulations, with one exception in 10 CFR 20.2005, provide no minimum level of licensed radioactive material that can be disposed of in a manner other than as radioactive waste or transferred to a licensed recipient.
SDP determination process: Is there an event or occurrence in the licensee's radiological material control program that is contrary to NRC regulations? If yes, the question is what is the dose impact (as calculated by the lic nsee) of the event? If the dose impact was not more than 0.005 mrem and there were not more than 5 of these events in the inspection period, then the SDP classification is Green. If the dose impact was greater than 0.005 mrem or there were more than 5 events that were not above 0.005 mrem in the inspection penod (may signify a rogrammatic breakdown), then the SDP classification is White. If the dose impact is greater than 0.1 mrem p(exceeds 10 CFR Part 20 public dose limit), the SDP classification is Yellow. If the dose impact was greater tha 0.5 rem, the SDP classification is Red.
Historically, these events have had calculated doses well below 0.001 mrem, thus, in most cases a Green significance determination is likely. However, if there were more than 5 events in the assessment period where licensed radioactive material was released, this may indicate a breakdown of the program.
Example:
The inspector reviewed survey records of material released from the restricted area of the plant. The records indicated that materials with no detectable licensed radioactive material were released for unrestricted use. During th3 inspection the licensee receives a call from another nuclear power plant that had received painting, equipment that was " free released" from the licensee. The radiation survey performed at that plant of the incoming painting equipment documented the presence of licensed radioactive material. The painting equipment was shipped directly from one plant to the other. The plant that received the contaminated painting equipment planned to return it to the first licensee (as a radioactive material shipment). The finding is that the licensee did not perform an adequate survey to prevent the inadvertent release of licensed radioactive material into an unrestricted area.
'To determine the significance requires a determination of the dose consequence to an individual from handling or being near the contaminated equipment. The licensee is responsible to evaluate the potential radiological hazard from the equipment. The significance determination will be based on the calculated dose for the event.
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I 06XX DRAFT A2-16 Issue Date: 05/07/99
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3 RYES Yellow Issue Date: 05/07/99 A2-21 DRAFT 06XX
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Attichmtnt 3 Physical Protection Significant Determination Process The safeguards risk assessment process recognizes that nonconformance issues have varying degrees of safety significance and in considering the safety significance of a nonconformance issue, it is appropnate to consider the technical significance (i.e., actual and potential consequences).
Once a nonconformance issue has been identified, the risk of radiological sabotage has to be determined.
The issue is evaluated to establish whether there is no/ low risk or more than low risk of
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radiological sabotage.
If there is no/ low risk, the issue is within the (licensee response band) and will be resolved via the licensee's corrective action program.
- 8. Examples of events within the (licensee response band):
a, A failure to make, maintain, or provide log entries in accordance with 10 CFR 73.71 (c) and (d)
- 9. A failure to conduct a proper search at the access point
- 10. A failure to control access such that an opportunity exists that could allow unauthorized and undetected access into the protected area but which was neither easily nor likely to be exploitable.
- 11. A failure to report acts of licensed operators or supervisors pursuant to 10 CFR 26.73.
- 12. A failure to perform an appropriate evaluation or background investigation so that information relevant to the access determination was not obtained or considered and, as a result a person was, granted access by the licensee who would probably not have been granted access d the required investigation or evaluation had been performed.
- 13. Isolated nonconformance with procedure requirements that are not indicative of a significant performance trend.
- 14. Protected / vital area barrier, alarm detection, and assessment nonconformance issues that do not impact an actual equipment performance (e.g., recording of test results, documentation for test sources, work request documentati
).
In the above examples, none of the events were aggravated by other factors (i.e., the events were not in conjunction with other safeguards failures; all were identified prior to any unauthorized entry into a vital area).
Repetitive events may be increased to a higher response band if the events can be considered a repetitive issue within the past 12 months.
The following nonconformance issues have been influenced by aggravating factors. These issues were discussed by headquarters and/ regional safeguards staff in order to yahdata the safeguards risk assessment process.
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- 1. A failure to protect safeguards information while information is outside the protected area and accessible to those not authorized access to the site.
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The Physical Security Plan contains details of the protection afforded the site. An unauthorized individual with Therefore, there is some risk involved with this event. Since the plan was unattended, prote malevolent intent could possibly exploit the safeguards systems and gain entry into a itis easily exploitable.
There were no other aggravated factors involved, that is, the plan was recovered, all security systems remained operable, and there was no unauthorized entry into the site. Since this was a single event involving safeguards information for the last 12 months, the issue is within the (licensee response band).
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- 11. The, entry into a vital area from outside the protected area by an unauthorized individual who damages safety equipment.
An unauthorized individual inside a vital area has exploited the protected area physical security systems and j
presents a risk to safety. The event has been aggravated by the failure of vital area barriers and intrusion Issue Date: 05/07/99 A2-23 DRAFT 06XX
detection system. Other safeguards mitigatin.g factors were " ineffective," that is, contingency response force failed to preclude unauthorized entry. Operational solutions were not successful and the calculated radiation dose exceeded the Commission guidelines established in 10 CFR 100. This event would be Category " Red."
However, if operation solutions were successful, this would fall within a (required regulatory response band).
Ill. A failure of protected area search equipment, thet is, metal / explosive detector /x-ray unit that results in the introduction to the protected area of firearms, explosives, or incendiary devices that could assist in radiological sabotage.
There is some risk associated with the event. If the contraband is available to unauthorized individuals with malevolent intent, it becomes easily exploitable. However, if the event is not aggravated by other factors, that is, it was detected and recovered before entry into a vital area, the risk of the event's contribution to radiological sabotage is low and could be dispositioned with the (licensee response band) if this was a single event involving unauthorized materials within the last 12 months.
Definitions:
Low Risk: A nonconformance activity has occurred that the licensee has determined presents no or low risk to the plant safety systems necessary to protect the public health and safety.
Predictable:
Based on the manner in which a program was being implemented. It was predictable that a violation would occur or that equipment, i.e., metal detectors, intrusion detection zones could be circumvented or d:feated without generating an alarm based on special knowledge obtained beforehand.
Exploitable:
If individuals are aware of equipment or system deficiencies and those deficiencies are not properly compensated for then those deficiencies are exploitable. They can be used to the greatest possible advantage by an individual (s) against the security organization.
Aggravating Factor: Any other factors that make the consequences of the event greater. Such as discovered d: graded protected and vital area barriers during an alarm assessment or a licensee drug screening facility not j
following good practices to ensure false specimens could not be substituted.
Operational Solutions: Intervention by control room personnel that would result in the safe shut down of the plint even if the contingency occurred and/or an adversary was able to render a piece of vital equipment inoperable or equipment configuration prevented an adversaiy from being successful in their attempt to endanger the public's health and safety..
06XX DRAFT A2-24 Issue Date: 05/07/99
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Appendix 3 Shutdown Activities inspection Findings Significance Determination Process To be Provided at a Later Time Issue Date: 05/07/99 A3-1 DRAFT 06XX
l Appendix 4 Fire Protection Program Finding Significance Determination Process To be Provided at a Later Time l
l 1
Issue Date: 05/07/99 A4-1 DRAFT 06XX
Appendix 5 SIGNIFICANCE DETERMINATION PROCESS AND ENFORCEMENT REVIEW PANEL This appendix provides guidance conceming a joint NRR/OE/RES/ Region review panel which has been cstablished to help ensure that the significance determination processes (SDP) described in appendices 1,2,3, e nd 4, which use a risk characterization, are implemented in a consistent manner. Consistency is important since
' SD.P provided the bases for the assessment and enforcement programs.
Th3 panel will include the following:
Th3 panel will be chaired by the Branch Chief or an attemate Section Chief from the inspection Program Branch 1
of NRR.
)
One member of the Operational Support Team from the Probabilistic Safety Assessment Branch, NRR.
J A management member or a designated attemate from the Office of Enforcement (OE).
f i
As designated by the panel chairman, there w31 be regional representation to include a DRP Branch Chief and a SRA from 1 of the 3 remaining regions not associated with the issue being reviewed.
~
~
)
Designated regional panel member and a SRA from the region associated with the issue being reviewed.
Other interested parties may attend by invitation of the panel members.
The regional panel member should normally be the pro ects section/ branch chief responsible for the site for which the SDP was conducted or another person designated ay regional manage, ment. Another regional section/ branch chief should be designated by regional management as the alternate.agional panel member.
G:neral Procedure:
for the purpose of reviewing all Phase 2 evaluations and to The panel will meet bi-weekly if necessary,ksignificantissuesbeingconsideredbeforetheyareissuedini independently discuss all proposed potential ns final form.
Th 3 designated regional panel member is responsible for bringing any proposed risk significant assessment inputs or violations to the attention of the panel in a timely manner so that the issuance of the inspection report is not unnecessarily delayed.
It is expected that all decisions regarding the assessment or enforcement actions will be made by consensus all members agreeing. If there is no consensus, the matter will be referred to the Director, Division of System Safety cnd Analysis, and the director of the Office of Enforcement for resolution.
The panel will also deal with any policy issues that are identified by any panel member or associated with the cetivities of the Operation Support Team from the Probabilistic Safety Assessment Branch, NRR.
Th] panel may also re uest program support from RES and other NRC groups to further development and use of the SDP. Additional, the panel may cause audits of the SDP to ensure appropriate guidance and training has been provided to the fi Id inspectors and their managers.
Tha panel shall meet, in person or by telephone conference call, on a schedule that is mutually agreed to by the panel members and that will not unnecessarily delay the issuance of the inspection report.
Th3 designated attemate may act for the member.
Others including, the reglonal inspector, resident inspector, project manager, etc., may be asked to attend tha meeting or provide input to the discussions.
Issue Date: 05/07/99 A5-1 DRAFT 06XX
l The panel shall maintain a record of all risk decisions results reviewed by the panel so they will be available i
for future comparison. Eventually, these will be used to develop a set of examples which could be added to the l
SDP assessment guidance.
The panel shall continue to review potential risk significant issues until NRC management agrees that such r; views are no longer needed. For the present, it is recommended that the panel plan on performing these reviews for the p)ilot plant elforts and the first year after the new reactor oversight process takes effect (i.e 1, 2001 or until adequate SDP guidance, with appropriate examples, is developed and provided to the regions.
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l l
l l
06XX DN A5-2 Issue Date: 05/07/99
{
l Management Directive 8.X
\\
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Attachment C 1
l l
J
U. S. NUCLEAR REGULATORY COMMISSION DIRECTIVE TRANSMITTAL TN:
To:
NRC Management Custodians
Subject:
Transmittal of Management Directive 8.X, " Reactor Oversight and Assessment Process (ROAP)"
Purpose:
Directive and Handbook 8.X have been issued to delineate the new Agency policy, objectives and implementation for the new reactor oversight and assessment process.
Office and Division of Origin:
Office of Nuclear Reactor Regulation Division of Inspection Program Management
Contact:
David Gamberoni,415-1144 James A.Isom,415-1109 Date Approved:
May XX,1999 Volume:
8 Licensee Oversight Program Directive:
8.X Reactor Oversight and Assessment Process Availability
. U.S. Government Printing Ofnce, (202) 512-2409
O REACTOR OVERSlGHT A~SD ASSESSMEST PROCESS (ROAP)
Directive 8.X
{
Contents:
I Poli cy..................................
1 O bj ecti ves..................................................................... ). 2 Organizational Responsibilities and Delegation of Authority............... 3 Executive Director for Operations (EDO)......................................... 3 Director, Office of Nuclear Reactor Regulations (NRR).............................. 3 Regional Administrators....................................................... 3 A pplica bili ty................................................................... 4 Ha nd book..................................................................... 4 Refere nces..................................................................... 4 1
i 1
)
i 1
Reactor Oversight and Assessment Process (ROAP)
Directive 8.X i
Policy (8.X-01)
It is the policy of the U.S. Nuclear Regulatory Commission to use the Reactor Oversight and Assessment Process (ROAP) to articulate the agency's observations and assessments regarding the licens2e's safety performance. The annual assessment letter communicates those observations and assessment to licensee management and the public.
Objectives (8.X-02)
To conduct an annual integrated assessment of licensee safety perfumance that focuses on the performance indicators and the safety significance of the NRC findings and conclusions.(021)
To provide a vehicle for meaningful dialogue with the licensee regarding its safety performance based on the infonnation gained from performance indicators and NRC findings and conclusions.(022)
To assist NRC management in maki pimely, predictable and scrutable decisions regarding allocation of NRC nesources used to oversee, inspect and assess licensee performance.(013)
To provide a method for infonaing the public of the NRC's assessment oflicensee performance.(024) l l
m Organizational Responsibilities and Delegations of Authority (8.X-03)
J Executive Director for Operations (EDO)
(031)
Overseas the activities described in this directive and handbook.
Director, Office of Nuclear Reactor Regulation (NRR)
(032)
Implements the requirements of this directive within NRR.(a)
{
e Monitors the ROAP process. Assesses the uniformity and adequacy of e
the implementation of the program.(b) e Evaluates and develops ROAP policy, criteria, and methodology.(c)
Regional Administrators (033) e Implements the requirements of this directive within their respective regions.(a) e Confer with the Director, NRR, and inform the Deputy Executive Director for Regulatory Programs when contemplating significant deviations from the requirements or guidelines of this directive.(b) e Develop and issue the annual assessment letter, which contains a concise i
assessment of licensee performance using performance indicators and NRC findings.(c)
Direct reallocation of regional inspection resources based on the Action e
Matrix and in accordance with the requirements set forth in Manual Chapter 2515.(d) e Establish a schedule and determine a site for a public meeting with the licensee, as appropriate, to ensure mutual understanding of the issues discussed in the annual assessment letter.(e) 2
Regional Administrators (033)(continued)
Suspend the end-of-year performance review for those plants which have e
been transferred to the Manual Chapter 0350 process.(f)
Assess the uniformity and adequacy of regionalimplementation of the e
ROAP Program. Provide to the Director, NRR, recommendations for improving the ROAP Program.(g)
Applicability (8.X-04)
The policy and guidance in this directive and handbook apply to all NRC employees.
Handbook (8.X-05)
Handbook 8.X addresses the major components of the ROAP Program.
References (8.x-06)
Inspection Manual Chapter 2515,06XX and 0350 1
Regulatory Assessment Performance Indicator Guideline (NEI 99-02(Draft) 3
REACTOR OVERSIGHT A'SD ASSESSME:ST PROCESS (ROAP)
Handbook 8.X o
l
Contents Part I General Guidance for the ROAP Program............................
1 Purpo se ( A)...............................................................
1 Assessment Process (B)........................................................
1 Assessment Frequency (C).....................................................
1 Assessment Areas (D).........................................................
1 Strategic Performance Areas, Cornerstones and Performance Indicators (1)........ 2 Risk-Informed Baseline Inspection Program and Significance Determination Process (2).................................... 2 Action Matrix (3).................................................... 4 Part II Implementation of the ROAP Program.............................. 5 Assessment Process (A)........................................................ 5 Continuous Reviews.................................................. 5 Quarterly Reviews................................................... 5 Mid.' Year Reviews..................................................... 5 End-of-Year / Agency Action Review....................................... 6 Commission Meeting................................................... 7 Meeting with Licensee (B)..................................................... 7 General ( l ).......................................................... 7 Meeting Preparation (2)................................................ 7 Conduct of Licensee Meeting (3)
....................................... 7 G l os sa ry..................................................................... 9 Exhibits 1
Regulatory Framework.........................................
10 2
Proce ss Table................................................. 1 1 3
Action Matrir..................................................
12 4
Window Summary of Performance Indicators and Inspection Results.....
13 5
Sample Letter: Inspection Addition Letter.........................
14 6
Sample I.etter: Six Month Inspection Look Ahead Letter...............
15
l 7
Sample Letter: Sample Annual Assessment Letter for Plants With All i
Assessment inputs (Performance Indicators (Pls) and Inspection Findings) G reen...............................................
17 8
Sample Letter: Sample Annual Assessment Letter for Plants With One or Two W hite Inputs. '............................................
19 9
Sample Letter: Sample Annual Assessment Letter for Plants With One Degraded Cornerstone........................................ 21 10 Sample Letter: Sample Annual Assessment Letter for Plants Which Requires Agency Review...................................... 23 l
L
Part I i
General Guidance for the ROAP Program Purpose (A)
The ROAP process is used to develop the NRC's conclusions regarding a licensee's safety performance. The annual assessment letter documents the NRC's observations and insights on a licensee's performance and communicates the results to the licensee and the public. It provides a vehicle for clear communication with licensee management that focuses plant performance based on performance indicators and NRC inspection results relative to safety risk perspectives. The NRC utilizes the ROAP when allocating NRC inspection resources at licensee facilities.
Assessment Process (B)
I The assessment process includes the fcllowing steps:(1)
I e
Periodic integration of data (quarterly, mid-year, and end-of-year) from performance indicators (PIs) and inspections by NRC staff and management at the appropriate levels to determine the level of safety performance being achieved by a particular plant. Enforcement actions taken should not be an input into the assessment process. However, the issue that resulted in the enforcement action will continue to be an input to assessment.(a)
Initiating scrutable, and predictable NRC actions based on the safety e
performance level of the plant.(b) e Communicating the NRC assessment and actions to the licensee and the members of the public(c) l The Regions will complete the quanerly, mid-year and end-of-year reviews of the plants.(2) l 1
The regional administrators (ras) or division directors (DDs), as appropriate, per exhibit 3 shall issue the end-of-year assessment letter.(3)
Appropriate level of regional management per exhibit 3 shall conduct a public meeting with the licensee's management to discuss the end-of-year assessment.(4) i Additional guidance regarding the ROAP process is provided in Part II of this handbook. Implementation procedures for the ROAP process will be contained in regional procedures. (5)
Evaluation Frequency (c)
The NRC will review and evaluate each power reactor licensee that possess an operating licensee annually with the following exceptions:(1)
)
1 End-of-year frequency and the scope of assessment may be adjusted for plants in extended shutdowns, extended outages, or decommissioning. In cach case, the regional administrator shall confer with the Director, NRR, and document the basis for the change to the assessments frequency.(a) j The ROAP process will be suspended for any plant that is shut down and requires authorization by the staff or the Commission to restart. As part of the restart review process, NRC will conduct a review consistent with requirements of Manual Chapter 0350.(b)
Each reactor site will have a separate assessment annually. For multiple unit sites, the regional administrator will assess performance of each unit and additionally, provide an overall assessment for the site.(2)
Assessment Areas (D)
The new regulatory oversight framework is designed to ensure that the NRC's overall safety mission is achieved through satisfactory plant performance in key strategic areas. Each strategic performance area, in turn, has a set of cornerstones (areas) that support the strategic performance area. The cornerstones are the essential attributes of a licensee's nuclear operations program that when these cornerstones objectives are fully met, they provide reasonable confidence that the safety goals are achieved. Performance must be maintained in the cornerstone areas to achieve the agency's strategic performance area goals and to meet the NRC's overall safety mission of ensuring adequate public health and safety as a result of civilian nuclear reactor operation.
Performance indicators and NRC inspections provide the data to assess the licensee's performance in the seven comerstone areas. Failure of licensees to submit accurate PIs in a timely manner may result in supplemental inspections.
2
Strategic Performance Areas, Cornerstones and Performance Indicators (1)
The three strategic performance areas are:
e Reactor Safety.
Radiation Safety.
Safeguards The seven comerstones in the three strategic areas and performance indicators used in these cornerstones are:
Reactor Safety.
Initiating Events. The performance indicators used for this cornerstone consists of unplanned scrams per 7000 critical hours; scrams with a loss of normal heat removal; and unplanned transients per 7000 critical hours.
Mitigating Systems. The performance indicators used for this cornerstone consists of selected safety systems for which their unavailability will have a notable adverse effect on reactor safety.
These systems include emergency power for all plants; high pressure coolant injection (HPCI), high pressure core spray (HPCS), reactor core isolation cooling (RCIC); and residual heat removal (RHR) for boiling water reactors (BWRs); and high pressure coolant safety injection (HPSI), auxiliary feedwater (AFW), and RHR for pressurized water reactors (PWRs).
e Barrier Integrity. The performance indicators for this cornerstone monitor the integrity of the fuel cladding, reactor coolant system (RCS) and containment. The indicators used are RCS specific activity, RCS identified leak rate and containment leakage, e
Emergency Preparedness. The performance indicators for this cornerstone are licensee's drill and exercise performance and the percentage of key emergency response organization personnel who have participated in a drill or exercise in the previous eight quarters.
Radiation Safety Public. The performance indicator for this cornerstone is the e
RETS/ODCM radiological effluent occurrences during the previous four quarters.
Occupational. The performance indicator for this cornerstone is the occupational exposure control effectiveness during the previous 12 quarters.
3
Safeguards Physical Protection. ihe performance indicators for this cornerstone e
are protected area security equipment performance index; personnel screening program performance; and fitness-for-duty fiFD) and personnel reliability program performance over the previous four quarters.
l l
Risk Informed Baseline Inspection Program and Significance Determination Process (2)
Risk-informed baseline inspection program (RIBIP) defines the scope of the l
NRC inspection program in the new ROAP. The RIBIP establishes the minimum regulatory interaction for all licensees. The inspection findings from the RIBIP are used together with the PIs to assess the overall licensee performance. RIBIP will cover those attributes of licensee performance not adequately covered by PIs RIBIP will also verify the accuracy of the PIs and l
provide for event response. The RIBIP is describe in NRC Inspection Manual Chapter 2515.
Significance determination process (SDP) is used to determine the risk significance of NRC inspection findings in the new ROAP. The SDP is used to evaluate the risk significance ofindividual inspection findings so that these findings can be compared and evaluated on a scale similar to the plant i
performance indicators. The SDP process is described in NRC Inspection Manual Chapter 06XX.
Action Matrix (3)
The action matrix details the range of licensee and NRC actions based on licensee performance. It also shows the expected levels of communications for a given action level. The action matrix was developed using the philosophy that the licensee should be allowed to address degrading performance issues first and the philosophy that NRC's action should be pursued based on the effectiveness of the licensee's corrective action and quality assurance programs. Action matrix is shown as Exhibit 3.
l l
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l 4
i
Part II Implementation of the ROAP Program Assessment Process (A)
The reactor oversight and assessment process is used to review the performance indicators and NRC inspection findings at different management levels and frequencies to assess the plant's safety performance. Continuos reviews are performed by inspectors; quarterly reviews are conducted by the Branch Chiefs in the Division of Reactor Projects' (DRP); mid-year and end-of-year myiews I
are conducted by regional management (division directors and regional administrators); and annual review to approve agency-level actions at the agency level.
i Continuous. A continuous review is conducted by the regional inspection staff.
Resident and regional inspectors shall maintain an awareness of plant performance. The inspectors will use the SDP to evaluate inspection findings, be aware of PIs that significantly cross thresholds, and be aware of non-SDP enforcement actions. Regional management will notify the licensee by letter if additional inspection activities are scheduled to occur within the current quarter.
Quarterly. DRP branch chief (BC) shall be responsible for conducting the quarterly performance review each quarter using branch resources. They shall complete their review within two weeks of receiving the PI date for the previous quarter. This is an informal data gathering and assessment activity, the primary purpose of which is to verify the accuracy of the quarterly date before releasing it to the public. The BC shall be responsible for ensuring that the PIs are obtained; that the PIinformation obtained are reasonably accurate; reviewing the PIs and the inspection findings contained in the plant issue matrix (PIM) to detect for any changes in performance trends. The BCs shall initiate and obtain approval for any additional NRC action in response to the change in performance. The BC will be responsible for notifying the licensee by letter only if additional inspection activities or other actions are to be performed based on the quarterly review.
Mid-year. The mid-year review uses the data compiled during the previous 12 months. The mid-year review is used primarily to plan and assign inspection activities.
5
The mid-year review meeting will be chaired by a DRP or Division of Reactor Safety (DRS) Division Directors (DD) or deputy DD and will be staffed by members of the DRP and DRS branches responsible for directing inspection resources. The mid-year review will be held at approximately the six-month point in the annual assessment cycle (within three weeks of the end of the second quarter).
The regions shall complete the mid-year review and notify the licensee by letter of planned inspection activities for the next six months and indicate the reason
- for planned inspections outside the normal baseline inspection, if any. The letter should be sent within one week of completing the mid-year review.
End-of Year / Agency Action Review. The end-of year review is intended to be the comprehensive review of plant performance by the Regions. The purpose of the end-of-year review is to conduct a comprehensive assessment of licensee performance using all PI and inspection data and to plan inspection activities for the next six months.
The end-of-year review meeting will be chaired by a DRP or DRS DD or deputy DD. A senior manager from DRP, DRS, and NRR will attend along with the DRP BC and inspectors and BCs with oversight of significant inspections at the site throughout the year and the Office ofInvestigations and Office of Enforcement. The results of the review will be an assessment letter for each plant. The letters will be presented to the RA for final approval. Results of the assessment will be compared to the action matrix to determine appropriate actions to consider. Performance warranting agency-level action will be forwarded to the agency action review.
Assessment letter shall contain:
l An overall statement regarding plant performance e-A statement (s) of any areas of concern.
An enumeration of any tripped assessment inputs from PIs or inspection findings.
l e
A statement of any action to be taken by the Agency An agency action review meeting is conducted by senior NRC managers, with the chairmanship by the Director of NRR, for plants needing additional review and approval of age.ncy-level actions, as defined by the action matrix. The review uses data compiled during the end-of-year review. The agency action review involves a collegial review of plants requiring additional oversight due to adverse performance, with senior regional management presenting assessment results and proposed NRC actions for selected plants. The review will take place (approximately two weeks) after the end-of-year review. The purpose of the review is to ensure a coordinated, balanced, and consistent agency response.
6
s Commission Meeting. The EDO will give the Commission an annual briefing to convey the assessment results for all plants, with focus on plants that required j
approval of agency-level actions,if any. The Commission will have negative i
consent on all assessment results and NRC r.ctions prior to their release. The Commission review should occur within eight weeks of the assessment cycle.
Meeting with Licensee (B)
General The ROAP annual meeting with the licensee will be scheduled within 30 days of the end of the assessment period to discuss the annual NRC assessment. The 30 day requirement may occasionally be exceeded to accommodate licensee management schedule conflicts. The meeting will be conducted onsite or in the vicinity of the site, if feasible, to foster accessibility and a more widespread understanding of the NRC's views by members of the public. Appropriate level of managers from the Regions as specified in the action matrix will conduct the public meeting. The Director, NRR, shall be informed in writing when a ROAP public meeting is not held.
Meeting Preparation The region shall notify those on distribution for the ROAP assessment letters of the meeting with the licensee.
The region shall notify the media and State and local government officials of the issuance of the ROAP assessment letter and of the meeting with the licensee once the letter has been released. Generally, at least I week's notice should be provided before the meeting.
The licensee will be encouraged to have the following management representatives participate in the meeting.
Senior corporate nuclear officer / manager Management officials responsible for the major functional areas Site managers Conduct of Licensee Meeting These meetings are intended to provide a forum for a candid discussion ofissues relating to the licensee's performance. Appropriate level of regional managers for the action matrix should discuss the assessment contained in the ROAP letter.
The licensee will be given the opportunity to respond at the meeting and to provide comments on the report in writing within 30 days after the meeting.
ROAP management meetings with the licensee will be public meetings, unless portions of the meetings involve discussion of the type of matters that are not 7
1 I
required to be publicly disclosed under Section 2.790 of Title 10 of the Code of l
Federal Regulations (10 CFR 2.790). For those portions, the meeting must be closed. Members of the public, the press, and Government officials should be treated as observers. ' Adequate notification of the ROAP meeting will be i
accomplisi.ed by the timely distribution to the Public Document Room, Local 1
Public Document Room of the letter scheduling the meeting with the licensee, and by posting notification of such meeting in the NRC website, with copies to j
the parties on the service list for the appropriate docket.
i
)
)
8
Glossary
'l Repetitive Degraded Cornerstone. A comerstone which is degraded (
2 white inputs or 1 yellow input) for five or more consecutive quarters (e.g. MS, MS, MS, MS, MS).
j 1
Multiple Degraded Cornerstone. Two or more cornerstones are degraded for five or more consecutive quarters.
Note: the degraded comerstones any vary throughout the period (e.g. E+MS, E+BI, E+MS, BI+MS, BI+MS)
MS = Mitigation Systems Cornerstone Degraded E = Initiating Events Cornerstone Degraded BI = Barrier Integrity Cornerstone Degraded i
9
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Exhibit 5 Sample inspection Addition Letter Licensee distribution designate Licensee name/ address
SUBJECT:
AdditionalInspections -(Plant Name)
(Use the following two sentences, as appropriate)
Our quarterly review of (plant name) identified that you have crossed the threshold (s) for the (insert performance indicator (s) threshold crossed) performance indicator (s). Additionally, we have identified significant inspection finding in the (name of cornerstone) area.
(Add additional information as appropriate)
This letter is to inform you that we will be planning supplemental inspection at your facility during the month of (month / year) to review (state what area you intend to review).
Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by), Chief Reactor Projects Branch Division of Reactor Projects Docket Nos. 50-ABC,50-XYZ Licensee Nos. NPF-0, NPF-0 cc:
Normal ce list 14
Exhibit 6 Sample Six Month inspection Look Ahead Letter Licensee distribution designate Licensee name/ address
SUBJECT:
Inspection Plan -(Plant Name)
On (date(s)), the NRC staff held a review of (plant name) to integrate performance information and to plan for inspection activities at your facility over the next (number) months. The purpose of this letter is to inform you of our plans for future inspection activities at your facility so that you will have an opportunity to prepare for these inspections and to provide us with feedback on
- any planned inspections which may conflict with your plant activities.
(Use one of the two paragraphs, as appropriate) 1.
We did not have identified any areas in which you crossed a performance threshold.
Therefore, we plan to conduct only baseline inspections at your facility over the next -
(number) months.
2.
(Use the following two sentences, as appronriate)
Our 6 month review of (plant name) identified that you have crossed the threshold (s) for j
the (insert performance indicator (s) threshold crossed) performance indicator (s).
i Additionally, we have identified significant inspection finding in the (name of cornerstone) area.
(Additional information, as appropriate)
Therefore, we plan to conduct additional (supplemental) inspections to better understand the causes contributing to your decline in performance.
During the next (number) months (iiie. 0/ peer to month / year), we plan to conduct the following inspections. These planned inspections include all planned inspections at your facility. includes all inspections which are scheduled to be accomplished during a specified week in the next (number) months. Enclosure 2 includes all other recurring inspections or inspections which we will perform as situations warrant. Additionally, these inspection activities are grouped by comerstone areas and contain the following information:
Planned Type of Number of Title / Program Area: Dates:
Inspection Activity:
Inspectors name of inspection mon / day / year-baseline or number of inspectors 4
mon / day / year supplemental 15
l l
If circumstances arise whibh cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by), Chief Reactor Projects Branch Division of Reactor Projects Docket Nos. 50-ABC,50-XYZ l
Enclosures:
1.
Scheduled inspection Plan 2.
Periodic inspection Plan 3.
Plant issue Matrix 4.
Window Summary of Performance Indicators and Inspection Results Cc.
l Normal cc list Distribution:
Normal distribution list plus Chief, NRR/DIPM/IIPB 16 l
l t
Exhibit 7 Sample Annual Assessment Letter for Plants With All Assessment Inputs (Performance Indicators (Pis) and inspection Findings) Green Licensee distribution designate Licensee name/ address
SUBJECT:
, Annual Assessment Letter -(Plant Name)
On (date(s)), the NRC staff completed the'end-of-year plant performance assessment of (plant name). The end-of-year review for (plant name) involved the participation of all technical divisions in evaluating performance indicators (Pis) and inspection results for the period (month / day / year to month / day / year). The purpose of this letter is to inform you of our assessment on your safety performance during this time period.
Overall, (plant name) operated in manner that preserved public health and safety. (Plant name) fully met all comerstone objectives.
As shown on enclosure 4, all performance indicators for the comerstones were in the licensee response band.
j Additionally, NRC inspection actmty and licensee self assessments did not identify any findings of safety significance in any of the comerstones.
Therefore, we plan to conduct only baseline inspections at your facility over the next (number) months.
During the next'(number) months (month / year to month / year), we plan to conduct the following inspections. These planned inspections include all planned inspections at your facility.
' Enclosure 1 includes all inspections which are scheduled to be accomplished during a specified
-)
week in the next (number) months. Enclosure 2 includes all other recurring inspections or j
- inspections which we will perform as situations warrant. Additionally, these inspection activities
- are grouped by comerstone areas and contain the following information:
Planned Type of Number of Title / Program Aros: Dates:
Inspection Activity:
inspectors name of inspection. mon / day / year-baseline or number of inspectors mon / day / year supplemental If circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by), Director Division of Reactor Projects, Region __
17 a
Docket Nos. 50-ABC,50-XYZ Licensee Nos. NPF-0, NPF-0
Enclosures:
1.
Scheduled Inspection Plan 2.
Periodic inspection Plan 1
3.
Plant issue Matrix 4.
Window Summary of Performance Indicators and inspection Results cc.
Normal ce list Distribution:
Normal distribution list plus Chief, NRR/DIPM/IIPB l
i 18
l l
Exhibit 8
. Sample Annual Assessment Letter for Plants With One or Two White inputs Licensee distribution designate l
Licensee name/ address l
SUBJECT:
Annual Assessment Letter -(Plant Name)
On (date(s)), the NRC staff completed the end-of-year plant performance assessment of (plant name). The end-of-year review for (plant name) involved the participation of all technical divisions in evaluating performance indicators (Pis) and inspection results for the period (month / day / year to month / day / year). The purpose of this letter is to inform you of our assessment on your safety performance during this time period.
Overall, (plant name) operated in manner that preserved public health and safety. (Plant name) fully met all cornerstone objectives.
[Use either one of the next two sentences, as appropriate, to discuss the Pis]
As shown on enclosure 4, all performance indicators for the cornerstones were in the licensee response band.
or As shown in enclosure 4, the performance indicators for the cornerstones were in the licensee l
response band with the following exceptions:
'(Provide PI(s) which crossed the threshold)
[Use either one of the next two sentences, as appropriate, to discuss NRC inspections)
Additionally, NRC inspections and licensee self assessments did not identify any findings of safety significance in any of the comerstones.
I or Additionally, NRC inspections identified / confirmed risk significant event (s) in (name of comerstone(s)).
[ Provide brief additional information about these events, as appropriate]
[lf these events have been reviewed by the licensee]
We have conducted additional inspections of your investigation into these events and we were j
satisfied with your review and proposed corrective actions.
[lf these events have not been reviewed by the licensee) 19
Therefore, we will perform additional inspections to review your investigations into these events and your proposed corrective actions.
During the next (number) months (month / year to month / year), we plan to conduct the following inspections. These planned inspections include all planned inspections at your facility. includes all inspections which are scheduled to be accomplished during a specified week in the next (number) months. Enclosure 2 includes all other recurring inspections or inspections which we will perform as situations warrant. Additionally, these inspection activities are grouped by cornerstone areas and contain the following information:
Planned Type of Number of Title /Procram Area: Dates:
Inspection Activity:
Inspectors name of inspection mon / day / year-baseline or number of inspectors mon / day / year supplemental if circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by), Director Division of Reactor Projects, Region._
Docket Nos. 50-ABC,50-XYZ Licensee Nos. NPF-0, NPF-0
Enclosures:
1.
Scheduled inspection Plan 2.
Periodic inspection Plan 3.
Plant issue Matrix 4.
Window Summary of Performance Indicators and Inspection Results i
cc.
Normal cc list Distribution:
I Normal distiibution list plus Chief, NRR/DIPM/HPB l
20
I Exhibit 9 Sample Annual Assessment Letter for Plants With One Degraded Cornerstone Licensee distribution designate Licensee name/ address
SUBJECT:
. Annual Assessment Letter -(Plant Name)
On (date(s)), the NRC staff completed the end-of-year plant performance assessment of (plant name). The end-of-year review for (plant name) involved the participation of all technical divisions in evaluating performance indicators (Pis) and inspection results for the period
- (month / day / year to month / day / year). The purpose of this letter is to inform you of our assessment on your safety performance during this time period.
Overall, (plant name) operated in manner that preserved public health and safety. (Plant name) met all comerstone objectives with minimal reduction in the safety margin.
(Cornerstone) was degraded.
[Use either one of the next two sentences, as appropriate, to discuss Pis]
As shown on enclosure 4, all performance indicators for the comerstones were in the licensee response band.
or As shown in enclosure 4, the performance indicators for the comerstones were in the licensee response band with the following exceptions:
(Provide Pls which crossed the threshold)
[Use either one of the next two sentences, as appropriate, to discuss NRC inspections)
Additionally, NRC inspections and licensee self assessments did not identify any findings of
. safety significanoe in any of the comerstones.
or Additionally, NRC inspections identified / confirmed risk significant event (s) in (name of comerstone(s)).
[ Provide brief additional information about these events, as appropriate]
[lf these events have been reviewed by the licensee]
We have conducted our own independent inspections of the events which resulted in a degraded comerstone. Further, we were satisfied with your self assessment, conducted with NRC oversight, of the causes contributing to the degraded comerstone.
21
[11 these events have not been reviewed by the licensee)
Therefore, you should conduct a self assessment into the causes for the degraded cornerstone.
Your self assessment efforts should be coordinated with my staff since it will require NRC oversight.' Additionally, we will conduct our own independent investigation into the causes for the degraded comerstone.
[Use either one of the next two sentences, as appropriate)
Because (comerstone) was degraded, this letter is to advise you that we believe a meeting with you would be appropriate. I will be contacting you to arrange for a mutually agreeable time and location for a meeting covering your decline in performance and your proposed actions to correct these deficiencies.
During the next (number) months (month / year to month / year), we plan to conduct the following inspections. These planned inspections include all planned inspections at your facility. includes all inspections which are scheduled to be accomplished during a specified week in the next (number) months. Enclosure 2 includes all other recurring inspections or inspections which we will perform as situations warrant. Additionally, these inspection activities are grouped by comerstone areas and contain the following information:
Planned Type of -
Number of Title / Program Area: Dates:
Inspection Activity:
Inspectors name of inspection mon / day / year-baseline or number of inspectors mon / day / year supplemental if circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by)
Regional Administrator, Region XX Docket Nos. 50-ABC,50-XYZ Licensee Nos. NPF-0, NPF-0
Enclosures:
1.
Scheduled Inspection Plan 2.
Periodic inspection Plan 3.
Plant issue Matrix 4.
Window Summary of Performance Indicators, d Inspection Results CC.
Normal cc list Distribution:
Normal distribution list plus Chief, NRR/DIPM/IIPB 22
1 l
Exhibit 10 1
Sample Annual Assessment Letter for Plants With Repetitive Degraded Comerstones, Multiple i
4 Degraded Comerstones, Multiple Yellow Inputs, or One Red input Licensee distribution designate
)
Licensee name/ address
SUBJECT:
- Annual Assessment Letter -(Plant Name) l On (date(s)),the NRC staff completed the end-of-year plant performance assessment of (plant
)
name). The end-of-year review for (plant name) involved the participation of all technical -
i divisions in evaluating performance indicators (Pis) and inspection results for the period (month / day / year to month / day / year). The purpose of this letter is to inform you of our assessment on your safety performance during this time period.
Overall, (plant name) operated in manner that preserved public health and safety. (Plant name) met all comerstone objectives with longstanding issues or significant reduction in safety margin.
[Use either one of the next two sentences, as appropriate, to discuss Pis]
As shown on enclosure 4, all performance indicators for the comerstones were in the licensee response band.
or As shown in enclosure 4, the performance indicators for the comerstones are in the licensee response band with the following exceptions:
(Provide Pls which crossed the threshoid)
[Use either one of the next two sentences, as appropriate, to discuss NRC inspections]
Additionally, NRC inspections and licensee self assessments did not identify any findings of safety significance in any of the comerstones.
or-Additionally, NRC inspections identified / confirmed risk significant event (s) in (name of comerstone(s))'
[ Provide brief additional information about these events, as appropriate]
Therefore, you should develop a performance improvement plan which will correct the deficiencies which are causing degradation of your comerstones. Your implementation of the performance improvement plan should be coordinated with my staff since it will require NRC oversight. Additionally, we will be conducting our own independent team investigation into the causes for the degraded cornerstone (s).
23
Because (comerstone(s)) was/were degraded, this letter is to advise you that we believe a meeting between the Executive Duty for Operation and your senior management would be appropriate. I will be contacting you to arrange for a mutually agreeable time and location for a meeting covering your decline in performance and your proposed actions to correct these deficiencies.
During the next (number) months (monthtfear to month / year), we plan to conduct the following inspections. These planned inspections include all planned inspections at your facility. includes all inspections which are scheduled to be accomplished during a specified week in the next (number) months. Enclosure 2 includes all other recurring inspections or inspections which we will perform as situations warrant. Additionally, these inspection activities are grouped by comerstone areas and contain the 'following information:
l Planned Type of Number of l
Title / Program Area: Dates:
Inspection Activity:
Inspectors name of inspection mon / day / year-baseline or number of inspectors mon / day / year supplemental if circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact (DRP Branch Chief) at (telephone number) with any questions you may have.
(Signed by)
Regional Administrator, Region XX Docket Nos. 50-ABC,50-XYZ Licensee Nos. NPF-0, NPF-0
Enclosures:
1.
Scheduled Inspection Plan 2.
Periodic inspection Plan 3.
Plant issue Matrix 4.
Window Summary of Performance Indicators and inspection Results cc.
Normal oc list Distribution:
Normal distribution list plus Chief, NRR/DIPM/IIPB 24
Rogulatory Assessment Performance Indicator Guideline i
i
,